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Enclosure 2 - Safety Evaluation Report for Review of Revision No. 22 of the Certificate of Compliance No. 9793 for the Model No. M-140 Transportation Package
	ML24253A258
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Site:	07109793
Issue date:	09/22/2024
From:	; Storage and Transportation Licensing Branch
To:	; US Dept of Energy, Naval Reactors
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ML24253A255	List: ML24253A256; ML24253A257; ML24253A258;
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

Safety Evaluation Report Docket No. 71-9793 Model No. M-140 Package Certificate of Compliance No. 9793 Revision No. 22 Summary

By letter dated March 29, 2023, (Agencywide Documents Access and Management System

[ADAMS] Accession No. ML23100A056), the U.S. Department of Energy, Division of Naval Reactors, (DOE-NR or the applicant) requested amendment to Certificate of Compliance (CoC)

No. 9793 for the Model No. M-140 package.

The U.S. Nuclear Regulatory Commission (NRC) staff performed its review of the M-140 package, provided in the safety analysis report for packaging (SARP or the application), utilizing the guidance provided in NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material: Final Report. Based on the statements and representations in the application, the analyses performed by the applicant demonstrate that the package provides adequate structural, thermal, containment, shielding, and criticality safety protection under normal conditions of transport (NCT) and hypothetical accident conditions (HAC), therefore the NRC staff concludes that the package meets the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 71.

1.0 General Information Evaluation

1.1 Packaging

The M-140 is a stainless steel package for transporting spent fuel. The overall package dimensions are 98 inches in diameter and 194 inches high. The package body is 14 inches thick with a closure head that is secured by 36 wedge assemblies located radially around the inside diameter. Penetrations in the closure head and body include an access port for fuel loading, vent and drain ports, water inlet and outlet penetrations, and a thermocouple penetration. The cask closure head and penetrations are sealed with plugs and double ethylene propylene O-ring seals. A stainless steel protective dome is positioned over the closure head. The cask body has 180 external vertical cooling fins, and a support ring is welded to these cooling fins. The support ring is bolted to a rail car mounting ring during transport. The fuel is positioned within an internals assembly. The internals assembly is composed of stacked spacer plates that have openings for the spent fuel modules. The maximum weight of the package, including contents, is 375,000 pounds.

1.2 Contents

The applicant revised the previously approved S3G-3 spent fuel contents by including an addendum that supplements existing analyses to account for the differences of the developmental material core (DMC) when compared to previously evaluated moored training ship (MTS) and shipboard S3G-3 cores.

1.3 Conclusion

Enclosure 2 The changes made were adequate and do not affect the continued ability of the package to meet the requirements in 10 CFR Part 71.

2.0 Structural Evaluation

The DOE-NRs application contained an addendum (Reference 1) to the S3G-3 spent fuel information in the M-140 SARP for the proposed DOE-NR CoC USA/9793/B(U)F-85, to allow shipment of the DMC spent fuel in the M-140 shipping container. The staff reviewed the application to verify that the structural performance of the package meets the requirements of 10 CFR Part 71.

2.1 Structural Design

The M-140 spent fuel shipping container (herein referred to as the M-140 container) is certified as a Type B package for shipment of fissile and highly radioactive material. The M-140 container has an S3G-3 internal component, which is designed to accommodate the S3G-3 and MTS fuel modules. The MTS modules were previously evaluated by the staff. The analysis and discussions presented in the core independent SARP (Reference 4) consist primarily of structural analyses of the container and the standard internals, and the containment aspects of the M-140 shipping container itself based on worst case assumptions about the cargo. The core specific analyses and evaluations are provided in the core specific SARP (e.g., Reference 3 for S3G-3 fuel modules with internal components) and associated addendums (e.g., Reference 2 for MTS fuel modules OR Reference 1 for DMC fuel modules), which supplement the core independent components evaluations provided in Reference 4. Thus, the core independent SARP (Reference 4) along with the core specific SARP (Reference 3) with any applicable addendum form a complete SARP for each cargo type.

The applicant stated that the DMC is installed in the modifications and additions to reactor facilities (MARF), which will be deactivated soon. The DMC is similar to an MTS core, but there are some fuel design differences as well as some unique modules that differ in geometry and cladding material, and some modules with minor geometric differences when compared to previously evaluated MTS (Reference 2) and shipboard S3G-3 cores (Reference 3), which were reviewed and accepted by the staff. As a result, the applicant submitted information on the differences and unique modules as an addendum to Reference 3. The addendum contains only the additional structural analyses to demonstrate the structural design adequacy of the DMC modules and any affected container and S3G-3 internal components under both NCT and HAC, as required by 10 CFR Part 71.

The staff notes that the application did not contain any additional structural analyses for the S3G-3 shipboard and container components if they were bounded by the previous structural analyses of the S3G-3 modules and internals provided in Reference 3 and the structural analyses of the container provided in Reference 4.

2.1.1 Description of Structures

The M-140 container is a right cylindrical stainless steel structure and is made of four principal structural components: (i) container body, (ii) closure head, (iii) protective dome, and (iv) S3G-3 internal, which holds the fuel modules in place. The major structural components of the S3G-3 internal in the M-140 container are the spacer plates, the top and bottom spring plates, the top plate assembly, and cell index plate, the support cylinder and the springs. The spent fuel claddings and weldments provide the primary containment boundary for the radioactive source material associated with the spent fuel cargo. The M-140 shipping container with the closure

head assembly provide containment for the crud, irradiated structural components and fuel modules, and is referred to as secondary containment. The applicant provided the dimensions and weights of the DMC fuel module in table B.1-2 of chapter B.1 of the application. In addition, the applicant provided the design drawings with detailed information in Appendix B.1.3.2 of the application to demonstrate compliance with 10 CFR 71.33(a) and 71.107(a).

The staff reviewed the structural descriptions and drawings for completeness and accuracy and finds that the application provides sufficient details to identify the package accurately and provide a sufficient basis for evaluation of the package.

2.1.2 Design Criteria

The design criteria used to evaluate the DMC package are the same design criteria used for the evaluations of the previous S3G-3 module package, as documented in section 2.1.2 of the SARPs (References 3 and 4). There are no new or revised design criteria in the application for the NCT and HAC evaluations.

The staff finds that the structural design criteria for the DMC S3G-3 package are acceptable because the design of the DMC package is similar to the design of the S3G-3 package (i.e., loads, weights, dimensions, configurations, materials, etc.), which was previously reviewed and accepted by the staff.

2.2 Weights and Centers of Gravity

The nominal weights and locations of the center of gravity (CG) of the DMC package components are provided in table B.2.2-1 of the application. These weights and CG are used for the structural evaluations to meet the NCT and HAC requirements of 10 CFR 71. The staff reviewed the information and determined that the applicant provided adequate information to describe the weights and determine the CG.

2.3 General Requirements for All Packages

2.3.1 Minimum Packaging Size

The smallest overall dimension of the DMC S3G-3 package is larger than the minimum requirement of 4.0 inches in 10 CFR 71.43(a).

The staff determined that the application meets the regulatory requirement of 10 CFR 71.43(a).

2.3.2 Tamper-Indicating Features

There is no design change proposed for the M-140 container in the application. The tamper-indicating features of the M-140 container, which were previously reviewed and accepted by the staff, are still valid for the DMC package.

The staff determined that the application meets the regulatory requirements of 10 CFR 71.43(b).

2.3.3 Positive Closures

There is no design change proposed for the M-140 container in the application. The positive closures of the M-140 container, which were previously reviewed and accepted by the staff, are still valid for the DMC package.

The staff determined that the application meets the regulatory requirements of 10 CFR 71.43(c).

2.4 Lifting and Tie-Down Standards for All Packages

The applicant stated that the DMC package meets the requirements of 10 CFR 71.45(a) and 71.45(b) for lifting devices and tie-down standards, respectively, because: (i) there is no design change proposed for the M-140 container, (ii) the locations of the center of gravity of the DMC package are within the acceptable range, and (iii) the total weight of the package remain bounded by that of the previously accepted S3G-3 package in Reference 3.

The staff concurs that the application meets the regulatory requirements of 10 CFR 71.45(a) and 71.45(b).

2.5 Normal Conditions of Transport

The applicant indicated that evaluations for the structural components of the shipping container and S3G-3 internals for the DMC shipment package under NCT are not repeated in the application because: (i) there are no physical changes to these structural components due to the DMC shipment, and (ii) the loading conditions of these structural components remain bounded (see explanation in the following paragraph) by the previous analyses of the shipping container contained in section 2.10.1.2 of Reference 4 and of the S3G-3 internal components contained in Appendix 2.10.1 of Reference 3.

The shipping container and the S3G-3 internals were previously analyzed for an internal design pressure of 75.0 psig (References 3 and 4), which bounds the maximum internal pressure of 13.8 psig due to the DMC shipment package. Also, the steady state temperatures for a DMC cargo shipment (table B.2.6-1) remain bounded by the steady state temperatures (SARP table 2.6.1) used in Appendix 2.10.1 of Reference 3 analyses for these components.

The staff agrees with the applicant's conclusion and finds it to be acceptable based on the above comparison to the previous evaluations contained in References 3 & 4, which were previously reviewed and accepted by the staff. The staff noted that the evaluation of the DMC fuel modules for NCT free drop conditions are enveloped and governed by HAC free drop evaluations, which are addressed under section 2.7 of this safety evaluation report (SER).

The staff determined that the DMC in the M-140 container satisfies the regulatory requirements of 10 CFR 71.71.

2.6 Hypothetical Accident Conditions

The applicant evaluated the DMC package for HAC free drop, thermal, and water immersion for fissile material as required by 10 CFR 71.73. However, the requirement for the HAC crush test is not applicable to a cask weighing more than 1100 pounds and therefore the crush test was not performed. The HAC puncture test remains bounding from section 2.7.2 of Reference 4, which causes local denting of the container and was considered for subsequent analyses.

The applicant analyzed the DMC fuel modules for a free drop using the closed-form solutions and a computer program, CRUSHTAB, which was used for the evaluations of the previous MTS and S3G-3 in the M-140 container in References 2 and 3 to demonstrate compliance with the regulatory requirements for HAC in 10 CFR 71.73. The CRUSHTAB program was developed to determine the impact deformations and the center-of-mass decelerations of the structural components in a shipping container impacting on an unyielding surface for free drops (flat end,

flat side and corner drops). The staff previously reviewed and accepted the adequacy of using the CRUSHTAB program (1990 version) for drop analyses in References 2 and 3.

Subsequently, the applicant has made changes to the program including the latest 2005 version and revised the Program Guide and Verification Manual for CRUSHTAB which is documented in Reference 5. The primary changes made to the program are summarized as follows:

added a subroutine for crush tests of new geometries that were not available in previous versions,

added zircaloy and concrete as a material option (also, added an option to ignore mushrooming; this option is needed primarily for concrete that does not mushroom like steel when it gets crushed under the impact),

added an option to ignore strain rate effects on material properties. It is added because turning off these effects conservatively bounds the crush depth, which is desirable for some materials which are not sensitive to strain rate,

added the capability to calculate and printout the elapsed time. This allows for generation of force and acceleration verses time curves. Previously only force and acceleration vs crush depth output data were available,

added the calculation and printout of maximum allowable crush depth for the straight edge, which clearly defined the allowable limits on crush depth and drop angle. This does not change how the crush area varies with crush depth.

The applicant verified and validated subroutines for the newly added shapes by comparing the CRUSHTAB output results to the hand calculations results for representative sample problems.

The applicant validated elapsed time results by comparing independent spreadsheet calculations with the program output; and verified correct implementation of the zircaloy and concrete strain rate properties by comparing program revisions independently performed by two individuals. The applicant validated remaining changes to the program by rerunning previously performed analyses using the updated program and comparing the results.

The staff reviewed the changes and related documentation and concluded that most of these changes primarily improve the program capability by including additional shapes and material and provide more flexibility in analyses and output results. The original methodology and the theoretical basis for the previous and new computation did not change and remained the same.

The applicant has successfully verified and validated the implementation of the algorithm for the newly added shapes and materials and other changes. Based on the above reviews, the staff concurs with the revised program validation and finds it acceptable for use in the analysis.

2.6.1 Free Drop

As required by 10 CFR 71.73(c)(1) a package needs to be demonstrated for structural adequacy by a free drop through a distance of 30 feet onto a flat, unyielding, horizontal surface in a position for which maximum damage is expected. In order to determine the orientation that produces the maximum damage, the applicant evaluated the DMC structural modules in the M-140 container for impact orientations in which the package strikes the impact surface. The applicant considered five drop configurations: (i) flat top drop, (ii) top corner drop, (iii) side drop, (iv) bottom corner drop, and (v) flat bottom drop.

The results of the drop analyses using the CRUSHTAB program for the structural components of the DMC structural modules are provided in Appendices B.2.10.2 through B.2.10.6 of the

application. The staff noted that the loading conditions during stable top corner and stable bottom corner drops are less severe for the DMC fuel modules components than for the flat top, flat bottom and side drops.

The results of the flat top drop, side drop and flat bottom drop analyses for the DMC fuel cells show that the fuel cluster in the fuel cells remain undamaged and the control rods remain intact and retained inside the cells containing control rods. Although some failure is predicted in some assemblies, the failure is shown to be acceptable. As a result, the control rods are predicted to withdraw somewhat from the cells. However, the calculated maximum withdrawal length remains bounding with respect to the length assumed in the criticality analyses of chapter B.6 of Reference 1.

As for the shipping container, the maximum force on the access plug under the flat top drop condition has increased for the minimum material property case (section B.2.10.2), which is greater than the design force evaluated in the previous analyses. This variation in the load is primarily attributed to the temperature associated with the minimum material properties case (450 °F for S3G-3/MTS vs. 200 °F for DMC) and other variations in the geometry and configuration of individual fuel module components. With the supplemental detailed evaluations, the applicant has shown that the increased force is acceptable for closure head retention system and access plug because no yielding occurs in any closure system components of the closure head access plug. The closure head retention system will retain the head and cargo during the flat top drop, with local yielding in wedges and the region of the container body flange ligaments between wedge assemblies, but will remain intact. These evaluations identify that ultimate strengths are not exceeded in the regions where yield strengths are exceeded (i.e., the wedges and flange ligaments in the wedge region of the container body flange). The staff noted that the associated plastic strains are very small (0.0307 and 0.036) and within the established design criteria limits of table 2.3-1 of Reference 4. Therefore, the increase in forces to the components does not impact their functions as described in the SARP.

Additionally, the results of the flat bottom drop analyses for the DMC fuel cells show that although yielding occurs, the energy absorbers prevent yielding in the fueled region, which indicate that the primary containment is maintained.

The results of the side drop analyses for the DMC fuel cells show that primary containment is maintained.

The staff reviewed the results of the CRUSHTAB analyses presented in Appendices B.2.10.2 through B.2.10.6 of the application. The staff also reviewed the calculated stresses and strains where applicable in the structural components and found that they are bounded by the design criteria stresses and strain limits established in section 2.1.2 of References 3 and 4. Based on its reviews and verifications, the staff confirmed the applicants findings.

The staff determined that the application satisfies the regulatory requirements of 10 CFR 71.73(c)(1).

2.6.2 Thermal

As required by 10 CFR Part 71.73(c)(4) exposure of the package to an average flame temperature of at least 1,475 °F must be for a period of 30 minutes.

The applicant performed thermal evaluations of the DMC package under HAC in chapter B.3 of the application. The following results are from a summary of the applicants thermal evaluations

in chapter 3 of the application: B2.7.3 states that primary containment of the DMC spent fuel will be maintained under both NCT and HAC fire accident.

The staff reviewed the statements presented by the applicant and found them acceptable.

Additional detailed reviews and safety evaluations by the staff on the applicants thermal evaluations are provided in chapter 3 of this SER.

The staff determined that the application satisfies regulatory requirements of 10 CFR 71.73(c)(4).

2.6.3 Immersion - Fissile Material

As required by 10 CFR Part 71.73(c)(5) for fissile material subject to 10 CFR Part 71.55, in those cases where water in-leakage has not been assumed for criticality analysis, it must be evaluated for immersion under a head of water of at least 3 feet in the attitude for which maximum leakage is expected.

The applicant stated that detailed criticality analyses for the DMC package under HAC are provided in chapter B.6 of the application, where the criticality analysis considered water in-leakage. Based on the results of the criticality analyses in chapter B.6, the applicant made a conclusion that the DMC package and its contents will remain subcritical under the worst-case condition of preferential flooding and control rod withdrawal. Therefore, the DMC package is acceptable under immersion conditions.

The staff reviewed the analyses and found it acceptable because it demonstrated subcriticality under the assumption that the DMC package allows in-leakage of water. The staffs detailed reviews and safety evaluations on the applicants criticality evaluations are provided in chapter 6 of this SER.

The staff determined that the application satisfies the regulatory requirements of 10 CFR 71.73(c)(5).

2.7 Fuel Cladding

The fuel cladding is relied upon to provide the primary containment boundary for nuclear fuel.

The applicant stated that the structural analyses provided in chapter 2 (Reference 3) and chapter B.2 of the application (Reference 1) and thermal analyses provided in chapter 3 (Reference 3) and chapter B.3 of the application (Reference 1) show that integrity of the fuel cladding is maintained under NCT and HAC.

The staff reviewed the results of the structural evaluations in chapters 2, B.2, 3 and B.3 of the SARPs (References 1 and 3) and found that the fuel cladding of the DMC in the M-140 container is not vulnerable to failure under 30 feet drop accident event. Additional safety evaluations with respect to the fuel cladding integrity are provided in chapter 3 of this SER.

2.8 Evaluation Findings

Based on a review of the statements and representations in the application, the staff concludes that the structural design has been adequately described and evaluated and that the DMC modules in the M-140 container transportation package have adequate structural integrity to meet the structural requirements of 10 CFR Part 71.

2.9 References

1. WAPD-REO(C)-1395, Addendum B - DMC Spent Fuel in the M-140 Safety Analysis Report for Packaging Addendum (U).

2. WAPD-REO(C)-1395, Addendum A - MTS Spent Fuel in the M-140 Safety Analysis Report for Packaging Addendum (U).

3. WAPD-REO(C)-1395, Rev. 5 - S3G-3 Spent Fuel in the M-140 Safety Analysis Report for Packaging (U).

4. WAPD-REO(C)-1600, Rev. 16 - Core Independent M-140 Safety Analysis Report for Packaging (U).

5. B-REO(C) -2975, Rev. 4 -- Program Guide and Verification Manual for CRUSHTAB.

3.0 Thermal Evaluation

The staff reviewed the M-140 SARP Thermal chapter to verify that the thermal performance of the package meets the requirements of 10 CFR Part 71.

3.1 Description of Thermal Design

The M-140 spent fuel shipping package includes a right circular cylindrical shell that is a 14-inch thick stainless steel forging with flange, 12-inch thick steel bottom plate, and 13-inch thick stainless steel closure head. Cooling fins are welded to the exterior shell of the package to aid in the passive cooling of the package. The closure head is held in place by a wedge closure system and is sealed using concentric O-rings against the exterior shell flange. The package bottom includes concentric stainless steel rings that act as an energy absorber. The loading and unloading operations of spent fuel are via an access opening in the closure head; the access opening is closed by a bolted shield plug. A stainless steel dome, which is attached to the packaging body, covers the closure head during transport. Although the M-140 packaging has penetrations for cooling water circulation, venting, and thermocouples, they are not used during transport and are sealed during shipment with plugs and double O-ring seals; therefore, there are no valves and no continuous venting from the packaging. The spent fuel content is held in place within the packaging by an internal assembly composed of stacked spacer plates.

According to the Core Independent M-140 SARP, the Type B fissile package is designed and constructed in accordance with military (MIL), American Society of Mechanical Engineers (ASME), and American National Standards Institute (ANSI) standards and the drawings contained in the SARP.

The M-140 package was previously reviewed and certified, including for the bounding contents described in the Core Independent M-140 SARP. Acceptance Tests and Maintenance chapter and the Operating Procedures chapters referred to the corresponding chapters in the previously reviewed Core-Independent M-140 SARP. This thermal evaluation is based on the thermal effects of the M-140 package with S3G-3 fuel cargo and the internals assembly. The M-140 package can be shipped when the decay heat levels are less than 7173 BTU/hr, which is evaluated in SARP chapter B.3.

3.2 Summary of Temperatures

According to the M-140 SARP Thermal chapter, the maximum temperatures of the accessible surface of an M-140 package with the S3G-3 at hot normal conditions, no insolation, and maximum decay heat were less than the 185 °F regulatory limit for exclusive use shipment described in 10 CFR 71.43(g). Predicted temperatures are provided in SARP chapter B.3 for both NCT and HAC.

3.3 Contents

SARP chapter B.3 stated that the additional M-140 shipping package contents seek in this revision includes DMC S3G-3 Fuel cargo. Decay heat levels are provided in chapter B.3, as stated in section 3.1 above.

3.4 Summary of Maximum Pressures

According to the S3G-3 fuel cargo the M-140 SARP Thermal chapter, package pressures at NCT and HAC were calculated based on the method described in the previously reviewed Core-Independent M-140 SARP. The Structural chapter indicated that the maximum normal condition internal pressure was less than the allowable pressure analyzed in the bounding Core Independent M-140 SARP.

3.5 Material Properties and Component Specifications

All material properties from chapter A.3 remain applicable except for the additional fuel material.

Temperature-dependent thermal properties of this fuel material are provided in SARP table B.3.2-1. The range covers both NCT and HAC. As noted earlier, according to the Core Independent M-140 SARP, the Type B package is designed and constructed in accordance with MIL, ASME, and ANSI standards and the drawings contained in the SARP.

3.6 Thermal Model Analyses

NCT Thermal model described in SARP section A.3.5 applies entirely to both NCT and HAC for this revision. The M-140 SARP Thermal chapter indicated that the finite element analysis (FEA)

ABAQUS code was used to generate a three-dimensional symmetric portion of the railcar structure, M-140 package (e.g., dome, closure head, M-140 vessel), S3G-3 spent fuel modules, and supporting structures. Although fins were not explicitly modeled, their effect on radiation heat transfer and on convection heat transfer by applying a fin effectiveness factor was analyzed using the methodology described in earlier M-140 amendments. The models convergence acceptance criteria were provided in the thermal chapter. Volumetric heating was applied to the spent fuel module to achieve the decay heat and axial profile. A bounding loading of spent fuel was assumed in the model. Spent fuel temperatures from the FEA code were calculated to ensure acceptable fuel performance criteria were met. The steady state thermal calculation for NCT hot conditions was based on 100 °F ambient temperature and the effect of insolation during NCT assumed the numerical values associated with flat and curved surfaces described in 10 CFR 71.71(c) but were conservatively applied for a 24-hour period (i.e., not for a 12-hour period). The SARP thermal chapter indicated that the models heat transfer boundary conditions applied to the packages outer surface finned area considered the increased convection and radiation heat transfer due to the fins, which, as noted earlier, were not explicitly modeled.

3.7 Thermal Evaluation Under Normal Conditions of Transport

SARP analysis during NCT shows that the maximum fuel and structural temperatures are acceptable for the M-140 shipment of the DMC core for a minimum hold time stated in the SARP. The accessible exterior surface temperature of the M-140 will not exceed 185 °F limit for an exclusive use shipment. In addition, the maximum temperatures and internal pressures of the package will not exceed the limits established in the SARP. Predicted temperatures show significant margin to fuel failure (as described in the SARP).

3.8 Thermal Evaluation Under Hypothetical Accident Conditions

Thermal expansion due to the fire is included in the model though modifications of respective gaps (during the entire transient analysis of the fire). Transient analyses are performed for sufficient duration to return to steady state temperature. SARP thermal evaluation shows that maximum fuel temperatures are acceptable for the M-140 shipment at the contained decay heat specified earlier in this SER. Thermal analysis results show predicted results provide sufficient margin for shipment of the DMC fuel in an M-140 container.

3.9 Evaluation Findings

Based on review of the statements and representations in the application, the staff concludes that the thermal design has been adequately described and evaluated, and that the thermal performance of the package with DMC fuel (S3G-3) contents meets the thermal requirements in 10 CFR Part 71.

4.0 Containment Evaluation

There were no changes to the package that affect the package containment system for the DMC S3G-3 fuel, therefore, for the DMC S3G-3 fuel, the M-140 package continues to meet the containment requirements in 10 CFR Part 71.

The M-140 spent fuel shipping container is a right circular cylinder that consists of a 14-inch thick stainless steel forged shell and flange, 12-inch thick steel bottom plate, and 13-inch thick stainless steel closure head; multiple pass welds are associated with the package and its containment boundary fabrication according to the Core-Independent M-140 SARP. Staff notes that cooling fins are welded to the exterior shell of the package to aid in the passive cooling of the package. The closure head is held in place by a wedge closure system and is sealed using concentric O-rings against the exterior shell flange. The package bottom includes concentric stainless steel rings that act as an energy absorber. The loading and unloading operations of spent fuel are via an access opening in the closure head; the access opening is closed by a bolted shield plug. A stainless steel dome, which is attached to the container body, covers the closure head during transport. Although the M-140 packaging has penetrations for cooling water circulation, venting, and thermocouples, they are not used during transport and are sealed during shipment with plugs and double O-ring seals; therefore, there are no valves and no continuous venting from the packaging. The spent fuel content (e.g., DMC S3G-3 fuel modules) is held in place within the shipping container by an internals assembly composed of stacked internal spacer plates. According to the Core-Independent M-140 SARP, the Type B package is designed and constructed in accordance with MIL specifications, ASME, and ANSI standards and the drawings contained in the SARP.

The M-140 shipping container was previously reviewed and certified, including for bounding content. The DMC S3G-3 M-140 SARP chapter B.7, Operating Procedures, and chapter B.8,

Acceptance Tests and Maintenance, referred to the corresponding chapters in the previously reviewed Core-Independent M-140 SARP. These chapters provide discussion related to leakage rate testing information. This containment evaluation is based on the SARP addendum for the DMC S3G-3 content consisting of the DMC S3G-3 fuel modules and the corresponding internals assembly. As noted in the DMC S3G-3 addendum M-140 SARP and the application letter, the DMC S3G-3 content has minor differences compared to the approved MTS S3G-3 content and is bounded by the previous evaluations associated with the Core-Independent M-140 SARP and the S3G-3 M-140 SARP.

4.1 Description of Content and Containment System

According to the DMC S3G-3 M-140 SARP chapter B.4, Containment, portions of the MTS S3G-3 M-140 SARP are still applicable, which the staff refers to in this SER. According to the MTS S3G-3 M-140 SARP chapter A.4, Containment, the activity of the Type B package is associated with the remaining fissionable material within the spent fuel (e.g., fission products, actinides), irradiated structural components, irradiated corrosion products (crud) that adhere to the surfaces of the fuel module and components, and the irradiated crud deposited on the interior surfaces of the M-140 packaging. The activity of the content was provided in the S3G-3 M-140 SARP. The MTS S3G-3 M-140 SARP portion of the Containment chapter, which is still applicable for the DMC S3G-3 M-140 SARP, noted that package activity is bounded by the content associated with the Core-Independent M-140 SARP. The CoC indicated a number of restrictions associated with DMC S3G-3 shipments of the M-140, such that shipments are restricted to a certain timeframe after reactor shutdown. It is noted that an increased number of days after reactor shutdown would reduce the contents activity, which would favorably affect shielding, containment, and thermal considerations.

According to the MTS S3G-3 M-140 SARP portion of the Containment chapter, the spent fuel cladding and weldments are the primary containment of the fuels remaining fissionable material; the cladding has no penetrations. The DMC S3G-3 M-140 SARP described two differences between the MTS fuel and the DMC fuel, which does not impact the containment analysis. The MTS M-140 SARP listed a number of acceptance tests associated with the fuel, including visual, ultrasonic and radiographic inspections. The M-140 packaging acts as the containment boundary for the radioactive content associated with the cladded fuel, irradiated structural components, irradiated crud that adheres to the fuel module surfaces and components, and the irradiated crud deposited on the interior surfaces of the M-140 package.

The M-140 packagings containment boundary consists of the bottom plate, cylindrical shell, bottom surface of closure head and associated welds as well as double O-ring seals, underside of top plate access plug, and the seals associated with transport container penetrations. The containment boundary extends to the inner O-rings. Details of the closure and containment boundary are provided in the Core-Independent M-140 SARP and drawings. Likewise, operating and testing aspects of the package, including leakage rate testing discussions, were provided in the previously reviewed Core-Independent M-140 SARP.

The previously reviewed Core-Independent M-140 SARP discussed the potential for radiolysis of the residual water within the M-140 shipping container and indicated that measurements have shown the hydrogen concentration from radiolysis would be less than 5 percent by volume.

4.2 Containment Under Normal Conditions of Transport

The DMC S3G-3 M-140 SARP Containment chapter table B.4.2-1 indicated that the NCT thermal analyses showed fuel temperature limit was met and that fuel cladding remains intact during NCT.

Likewise, the DMC S3G-3 M-140 SARP Containment chapter table B.4.2-1 stated that the thermal and structural analyses showed the integrity of the M-140 shipping container containment boundary was maintained for NCT and, according to the Core-Independent M-140 SARP, that seal temperatures were within allowable hot and cold temperature range. The DMC S3G-3 M-140 SARP Containment chapter table B.4.2-1 indicated that the internal pressure of the M-140 package with DMC S3G-3 content was bounded by the internal pressure associated with the previously reviewed bounding S3G-3 content analysis and the Core-Independent M-140 analysis.

The DMC S3G-3 M-140 SARP Containment chapter indicated that seal leakage rate tests per ANSI N14.5-2014 (Reference 1) requirements are performed when shipping MTS S3G-3 content to ensure the 10 CFR 71.51(a)(1) regulatory requirement of no dispersal at NCT is met (as demonstrated to a sensitivity of 1E-6 A2/hr); details of the tests were referenced in the Core-Independent M-140 SARP. The DMC S3G-3 M-140 SARP Containment chapter described that because there is less crud for the DMC S3G-3 content, as described in the DMC S3G-3 M-140 SARP and also described in section 4.3 of this SER, as well as a lower maximum internal pressure than the value assumed for NCT in the Core-Independent M-140 SARP, the reference air leakage rate is higher than the bounding spent fuel evaluated in the Core-Independent M-140 SARP, such that the potential release of radioactive material is bounded by the evaluation in the Core-Independent M-140 SARP.

Based on the fuel cladding integrity and the M-140 containment boundary integrity being maintained during NCT as described above, and the inclusion of leakage rate tests described in the previously reviewed Core-Independent M-140 SARP, staff finds that 10 CFR 71.43(f) and 71.51(a)(1) are met.

4.3 Containment Under Hypothetical Accident Conditions

As mentioned in the MTS S3G-3 M-140 SARP portion of the Containment chapter, which is still applicable for the DMC S3G-3 M-140 SARP, the fuel cladding did not fail due to structural-related HAC tests and the high temperatures that can occur as a result of the thermal fire accident condition. Therefore, the integrity of the fuel cladding would be maintained.

However, the MTS S3G-3 M-140 SARP portion of the Containment chapter, which is still applicable for the DMC S3G-3 M-140 SARP, indicated that the structural analyses presented in the Core-Independent M-140 SARP showed that as a result of the HAC tests, a slight gap may form at the M-140 shipping container closure head seating surface. Therefore, the HAC containment analysis for the MTS S3G-3 content was based on demonstrating that certain aspects associated with the MTS S3G-3 content, including the amount of crud and shipping environments, were bounded by the content evaluated in the Core-Independent M-140 SARP that met the containment criteria in 10 CFR 71.51(a)(2) after HAC. This same concept applies to the DMC S3G-3 content. For example, the DMC S3G-3 M-140 SARP Containment chapter indicated that the DMC S3G-3 fuel surface area is not greater than the MTS S3G-3 fuel surface area, and the MTS S3G-3 fuel surface area is less than the bounding content surface area, such that there would also be less crud for the DMC S3G-3 content compared to the bounding content. Likewise, it was shown that the MTS S3G-3 content and bounding content have consistent coolant chemistry, which is still applicable for the DMC S3G-3 M-140 SARP, and the DMC S3G-3 content has longer times between shutdown and shipment, thus indicating the analysis in the Core-Independent M-140 SARP remained valid. In addition, it was stated that the DMC S3G-3 content and the bounding content have similar shipping environments (e.g.,

temperature and pressure) such that the Core-Independent M-140 SARP results are valid for

the DMC S3G-3 content. Based on the above, the DMC S3G-3 M-140 SARP Containment chapter indicated that because 10 CFR 71.51(a)(2) is met for the bounding content previously reviewed, the regulation also would be met for the DMC S3G-3 content within the M-140 package during HAC.

Based on the above, staff finds that the DMC S3G-3 content within the M-140 package would satisfy 10 CFR 71.51(a)(2).

4.4 Evaluation Findings

Based on a review of the statements and representations in the application, the staff concludes that the containment design has been adequately described and evaluated, and that the package design meets the containment requirements of 10 CFR Part 71.

4.5 References

1. American National Standards Institute, ANSI N14.5-2014, American National Standard for Radioactive Materials - Leakage Tests on Packages for Shipment, New York, NY.

5.0 Shielding Evaluation

The applicant requested an amendment to the certificate for the M-140 package. The M-140 container is designed to accommodate spent fuel modules from several different cores. The container is the same for all spent fuel but has different internal assemblies to adapt to different spent fuels. The S3G-3 internals are designed to accommodate S3G-3 spent fuel.

The major change related to the shielding evaluation is for the DMC. The objective of this review is to verify that the M-140 shipping container meets the external radiation requirements of 10 CFR Part 71 under NCT and HAC for the DMC contents.

The DMC fuel is the same as S3G-3 fuel but with a modification. The changes are to the non-fuel structural components of some of the fuel cells and some of the internal features.

The weight of DMC fuel elements and associated shipping components are displayed in SARP tables B.1-2 and B.2.2-1. The weight of the fuel shown in section 1.1.1.9.4 are applicable to all M-140 shipping, including the DMC addendum.

5.1 Content of Packing

The DMC consists of reactor modules. The DMC modules are similar to S3G-3 with exception of some cells. Certain cells have cladding that is different from S3G-3 fuel assemblies. The other cells are DMC driver cells.

The clad test cell (CTC) is like the DMC cell except for the cladding material. The applicants safety evaluation is based on an evaluation of worst case S3G-3 and DMC spent fuel modules.

The worst case identified by the applicant is reactive S3G-3 modules at the most reactive credible time when removed from core. The DMC power density is bounded by the S3G-3 core design, and the minimum hold time of spent fuel from reactor shutdown to shipment in the M-140 is 3 years.

The quantity of the radionuclides in section 1.2.3.1 is applicable to all shipments since the quantity of S3G-3 cooled for 120 days after shutdown is conservative when compared to the minimum 3 years after shutdown for DMC fuels.

5.2 Description of Shielding Design

5.2.1 Design Features

The M-140 package consists of DMC fuel with a modification. The M-140 is normally shipped via rail car inside a well ring and is held in place by a support ring integral to the car. Items to be shipped in the package include spent fuel modules and assembly internals.

5.2.2 Summary Tables of Maximum Radiation Levels

Maximum radiation levels allowed under 10 CFR Part 71 on contact with the package surface and at 2 m under NCT are 200 mrem/hr and 10 mrem/hr, respectively. Maximum radiation levels on contact with the package surface under HAC are 1000 mrem/hr. Radiation levels are summarized in tables in chapter 5 of the SARP for NCT and HAC, respectively, and are below the limits listed above.

5.3 Radiation Source

The applicant indicated that the shielding analysis conservatively determined the source strength by basing it on the analysis provided in the SARP. The DMC fuel modules are almost the same as standard S3G-3 core with the same S3G standard mechanical assemblies. The source geometry dimension display in the drawing 1329J07. The applicant used an effective radius (as specified in the SARP) for sectional area evaluation. The staff reviewed the radiation source and finds this acceptable as it results in the most limiting radiation source.

5.4 Gamma Source

The applicant stated that they modeled the gamma source with axial variations, including segments below and above the fuel region. The applicant based the fission product gamma distribution on the maximum power generated by the most depleted fuel assembly at any time in core life. Total gamma strength distributions are provided in chapter 5 of the SARP.

The DMC fuel is assumed to be in shut down 120 days, which is the same as the shut down for S3G-3 fuel evaluated in the chapter 5 of the SARP.

Staff observed that the applicants model includes assumptions that overestimate the contribution of subcritical multiplication to the gamma source. When determining the gamma source from component activation, the applicant applied administrative factors to the thermal neutron reaction rate, modeled materials with the maximum amount of activation impurities, and assumed a neutron fall-off rate that maximized component activation. The applicant includes the gamma rays induced by the absorption of neutrons by the fission products in addition to those directly from fission product decay.

Staff finds the applicants assumptions are consistent with the previous fuel evaluated in the SARP and was performed according to the guidance in NUREG-2216, Standard Review Plan for Transportation Packages for Spent Nuclear Fuel and Radioactive Material and NUREG/CR-6802, Recommendations for Shielding Evaluations for Transport and Storage Packages. Based on this amendment to the SARP and prior staff review of the chapter 5 for

S3G-3 fuel, staff concludes that DMC gamma source is bounded with previous gamma source in chapter 5 of SARP.

5.5 Neutron Source

The applicant chose an isotope energy spectrum to represent photoneutrons and transuranic neutrons and chose a different isotope spectrum for subcritical multiplication. The applicants analysis also accounted for certain decay reactions and the effect of subcritical multiplication in determining the neutron source strength. The applicant used the irradiation histories for photoneutron and transuranic neutron sources that maximized each individually as stated in SARP chapter 5 to evaluate the DMC neutron source. Based on the analysis in the SARP, the staff finds that the neuron source in chapter 5 bounds the neutron source of the DMC fuel.

5.6 Shielding Evaluation

The staff found the applicant described and/or provided drawings of sufficient detail for NRC staff to confirm the applicants analysis for both NCT and HAC. Since both gamma source and neutron source are bounded with the previous evaluation in SARP chapter 5, the applicant didnt evaluate any shielding calculations and concluded that it is bounded with previous evaluation, and therefore meets the regulations of 10 CFR 71 for both NCT and HAC.

5.7 Evaluation Findings

Based on staff review of the methods, analyses, information presented in the application, and prior staff review, for the reasons discussed above, staff finds with reasonable assurance that the shielding requirements of 10 CFR Part 71 will be met with the proposed DMC contents and packaging design.

6.0 Criticality Evaluation

Staff reviewed the proposed changes to the S3G-3 spent fuel in the M-140 shipping container, which included minor modifications to the upper fuel module structure as well as the addition of the DMC. The DMC is similar to the previously approved core designs and was evaluated as a new content to ensure compliance with 10 CFR Part 71.

The applicant evaluated the M-140 package loaded with DMC modules using conservative and optimum conditions that maximized the reactivity of the modules using an approved three-dimensional Monte Carlo methodology. NCT and HAC were evaluated for a single package and arrays of packages and found that the calculated keff to be below 0.95 in all instances, in compliance with the regulations of 10 CFR Part 71. The calculated Criticality Safety Index of S3G-3 fuel was confirmed to be 100.

6.1 Evaluation Findings

Staff evaluated the criticality calculation methodology, computer code utilized, the cross-sectional data sets used, as well as the conservativisms used to maximize the reactivity of the packaged DMC modules. Since the resulting keffs for the system under both NCT and HAC were confirmed through the applicants analysis to be less than 0.95, staff concludes that the M-140 package containing DMC modules under the assumptions utilized by the applicant continues to meet the criticality safety requirements of 10 CFR Part 71.

7.0 Materials Evaluation

The staff reviewed the application (Reference 1), which is an addendum to the SAR, Revision 16 for DOE-NR CoC USA/9793/B(U)F-85 (DOE-NR) to ship the DMC cargo in the M-140 spent fuel shipping container, to verify that the material performance of the package meets the requirements of 10 CFR Part 71. Only the sections of the materials evaluation that changed from the previous SARPs (References 2 and 3) will be discussed below.

7.1 Materials of Construction

The applicant submitted an application, including an addendum to Reference 2, to demonstrate the material adequacy of the previously approved M-140 spent fuel shipping container (herein referred to as the M-140 container), and the previously approved S3G-3 internals (designed to accommodate S3G-3 fuel modules), to transport the DMC spent fuel modules. No changes in the materials of construction were made to either the M-140 container or S3G-3 internals. The DMC spent fuel modules are similar to the previously evaluated MTS fuel modules in the previously approved package described in Reference 4 with some fuel design differences, with the exception of a portion of the modules that have minor geometric differences, and a portion of modules that are unique. For some of the DMC spent fuel modules, these design differences involve new variations of cladding materials. Per the above discussion, the staff finds that the applicants description of the materials of construction to be acceptable.

7.2 Material Properties

The applicant did not make any changes to the mechanical and thermal properties of the materials used in the structural and thermal analysis for the M-140 container or S3G-3 internals.

The applicant did provide material properties for the new variations of cladding materials. The staff reviewed these material properties provided in table B.2.3-1 and verified that they are consistent with military handbook values. The applicant also provided corrected lower values of ultimate strength for a stainless steel used in the DMC cargo. The staff reviewed these updated mechanical properties provided in table B.2.3-1 and verified that they are consistent with military handbook values. Therefore, the staff finds the material properties used in the applicants structural and thermal analysis to be acceptable.

7.3 Corrosion Resistance and Content Reactions

The staff reviewed the revision changes and verified that they do not introduce any adverse corrosive or other reactions that were not previously considered in the staffs prior review of the M-140 Shipping Container. The materials of construction and the service environments are bounded by those that were previously evaluated in the CoC. Therefore, the staff finds the applicants evaluation of corrosion resistance and potential adverse reactions to be acceptable.

7.4 Spent Fuel

As stated earlier, the DMC spent fuel modules are like the previously evaluated MTS fuel modules in the previously approved package described in Reference 4 with some fuel design differences, with the exception of a portion of the modules that have minor geometric differences, and a smaller portion of modules that are unique.

The applicant provided a determination of fracture toughness for one of the materials of construction of the fuel hardware to account for a reduction of toughness due to thermal and radiation embrittlement. The applicant showed that this reduction in fracture toughness could lead to possible damaged structural members in the fuel hardware in an accident. The applicant demonstrated that this damage to structural members would not compromise the ability of the

fuel assembly to perform its intended functions, nor would it compromise its ability to meet fuel-specific and package-related regulations.

The applicant provided a thermal analysis in Appendix B.3 to evaluate the fuel performance during NCT and HAC. The staff reviewed this analysis and verified that adequate margin is provided to the maximum fuel temperatures that could result in fuel cladding failure.

Therefore, the staff finds that applicants description of the package contents to be acceptable and the mechanical properties of the fuel modules are adequate to ensure that the SNF remains in the analyzed configuration under NCT and HAC.

7.5 Evaluation Findings

F7.1 The applicant has met the requirements in 10 CFR 71.33. The applicant described the materials used in the transportation package in sufficient detail to support the staffs evaluation.

F7.2 The applicant has met the requirements in 10 CFR 71.43(f) and 10 CFR 71.51(a). The applicant demonstrated effective materials performance of packaging components under NCT and HAC.

F7.3 The applicant has met the requirements in 10 CFR 71.55(d)(2). The applicant has demonstrated that the package will be designed and constructed such that the analyzed geometric form of its contents will not be substantially altered and there will be no loss or dispersal of the contents under the tests for NCT.

The staff concludes that the Addendum to Reference 2 adequately considers material properties and material quality controls such that the design is in compliance with 10 CFR Part 71. This finding is reached on the basis of a review that considered the regulation itself, appropriate regulatory guides, applicable codes and standards, and accepted engineering practices.

7.5 References

1. WAPD-REO(C)-1395 - DMC Spent Fuel in the M-140 Safety Analysis Report for Packaging Addendum (U).

2. WAPD-REO(C)-1395, Rev. 5 - S3G-3 Spent Fuel in the M-140 Safety Analysis Report for Packaging (U).

3. WAPD-REO(C)-1600 - Core Independent M-140 Safety Analysis Report for Packaging.

4. WAPD-REO(C)-1395, Addendum A - MTS Spent Fuel in the M-140 Safety Analysis Report for Packaging Addendum (U).

8.0 Operating Procedures

There were no changes to the package that affect the operating procedures, therefore the M-140 package continues to meet the containment requirements in 10 CFR Part 71.

9.0 Acceptance Tests and Maintenance Program

There were no changes to the package that affect the acceptance tests or maintenance program therefore the M-140 package continues to meet the containment requirements in 10 CFR Part 71.

CONDITIONS

In addition to minor editorial changes, the following changes have been made to the certificate:

Condition 5b.(2)(i) was revised to include the decay heat per package for shipboard and MTS cores.

Condition 6(b) was revised to clarify the minimum fuel cooling time for shipboard and MTS cores versus the DMC.

Condition 8(b) was revised to reflect the correct minimum fuel cooling time for middle-of-life shipboard cores.

Conditions 7, 9, and 10 were revised at the request of the applicant due to reasons of sensitivity.

The REFERENCES section was revised to include the date of the application.

CONCLUSION

These changes do not affect the ability of the package to meet the requirements of 10 CFR Part 71.

Issued with Certificate of Compliance No. 9793, Revision No. 22.

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