ML23214A305

Enclosure 2 (Redacted) - Safety Evaluation Report-Revalidation Recommendation for the French Approval Certificate Number (No.) F/348/AF-96, Revision Gw, Model No. FCC-4 Transport Package (Docket No. 71-3097)

ML23214A305
Person / Time
Site:	07103097
Issue date:	08/08/2023
From:	Garcia-Santos N Division of Fuel Management
To:	Boyle R US Dept of Transportation, Radioactive Materials Branch
Shared Package
ML23209A340:ML23179A268	List: ML23209A635 ML23214A305
References
A33010, EPID L-2022-DOT-0008
Download: ML23214A305 (27)

Text

OFFICIAL USE ONLY - PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Model No. FCC-4 Package French Competent Authority Certificate No. F/348/AF-96 Revision Gw Docket No. 71-3097 Enclosure 2 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION Table of Contents Page

SUMMARY

....................................................................................................................................1 1.0 GENERAL INFORMATION REVIEW .................................................................................2 1.1 Package Description ..............................................................................................2 1.1.1 Packaging ..................................................................................................2 1.1.2 Contents.....................................................................................................2 1.2 Drawings ................................................................................................................3 1.3 Evaluation Findings................................................................................................3 2.0 STRUCTURAL EVALUATION............................................................................................3 2.1 General Considerations .........................................................................................3 2.2 Structural Analysis under Routine Conditions of Transport ...................................3 2.3 Structural Analysis Under Normal Conditions of Transport ...........................4 2.4 Structural Analysis Under Accident Conditions of Transport.........................5 2.4.1 Mechanical Property Changes of Fuel Assembly Materials.......................5 2.4.2 Mechanical Property Changes of Shell and Screw Steels .........................7 2.4.3 Numerical Analysis of the FCC-4 Package for a 9-m Vertical Drop...........7 2.5 Evaluation Findings................................................................................................7 3.0 THERMAL EVALUATION...................................................................................................8 3.1 Description of Thermal Design...............................................................................8 3.2 Material Properties and Component Specifications ...............................................8 3.2.1 Material Properties .....................................................................................8 3.2.2 Component Specifications .........................................................................8 3.3 General Considerations .........................................................................................8 3.4 Thermal Evaluation under Normal Conditions of Transport...................................9 3.4.1 Heat and Cold ............................................................................................9 3.5 Thermal Evaluation under Accident Conditions of Transport ................................9 3.5.1 Initial Conditions.........................................................................................9 3.5.2 Fire Test Conditions ...................................................................................9 3.5.3 Maximum Temperatures and Pressure......................................................9 3.5.4 Maximum Thermal Stresses ....................................................................10 3.5.5 Analyses Details.......................................................................................10 3.6 Evaluation Findings .............................................................................................10 4.0 CONTAINMENT EVALUATION .......................................................................................10 4.1 Description of the Containment System ..............................................................10 4.2 General Considerations .......................................................................................11 4.2.1 Classification as a Type A-Fissile Package .............................................11 4.2.2 Gas Generation........................................................................................12 4.2.3 Radiation Shield .......................................................................................12 4.3 Containment Evaluation under Accident Conditions of Transport .......................12 5.0 MATERIALS EVALUATION .............................................................................................13 5.1 Drawings ..............................................................................................................15 5.2 Materials Standards .............................................................................................15 5.3 Weld Design and Inspection ................................................................................17 5.4 Mechanical Properties .........................................................................................17 5.4.1 Low-Alloy and Carbon Steels...................................................................17 5.4.2 Stainless Steels........................................................................................18 5.4.3 Wood (Impact Limiter)..............................................................................18 5.5 Thermal Properties of Materials...........................................................................18 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.6 Criticality Control Materials ..................................................................................18 5.7 Corrosion and Chemical Reactions .....................................................................19 5.8 Content Integrity: Fresh (Unirradiated) Fuel Cladding .........................................20 5.9 Evaluation Findings .............................................................................................21 6.0 CRITICALITY SAFETY EVALUATION.............................................................................21 6.1 Design Description...............................................................................................21 6.2 Evaluation Findings .............................................................................................22 7.0 QUALITY ASSURANCE ...................................................................................................22 7.1 Evaluation of the Quality Assurance Program .....................................................22 7.2 Evaluation Findings .............................................................................................23 8.0 MAINTENANCE PROGRAM............................................................................................23 CONDITIONS ..............................................................................................................................23 CONCLUSION ............................................................................................................................23 REFERENCES ............................................................................................................................23 List of Tables Table 1. [Table Withheld per 10 CFR 2.390] ..............................................................................4 Table 2. [Table Withheld per 10 CFR 2.390] ..............................................................................5 Table 3. [Table Withheld per 10 CFR 2.390] ..............................................................................6 Table 4. Material properties design criteria for zirconium alloys. ..................................................6 Table 5. Documents related to the thermal performance of the Model No. FCC-4 package. .......9 Table 6. Standards used in the design of steel materials used in the design of the FCC-4........16 Table 7. Standards related to bolting materials used in the design and fabrication of the FCC-4.

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION REPORT Model No. FCC-4 Package French Competent Authority Certificate No. F/348/AF-96 Revision Gw Docket No. 71-3097

SUMMARY

By letter dated August 3, 2022 (DOT, 2022), as supplemented on January 26, 2023 (DOT, 2023a), and May 31, 2023 (DOT, 2023b), the U.S. Department of Transportation (DOT) requested that the U.S. Nuclear Regulatory Commission (NRC) staff (the staff thereafter) perform a review of the French Competent Authority Certificate (FCAC) (No.) F/348/AF-96, revision Gw, Model No. FCC-4 transport package (FCC-4 or FCC-4 package thereafter), and make a recommendation concerning the revalidation of the package for import and export use.

Specifically, the DOT requested NRC review changes such as:

1) Add zirconium alloy cladding with 30 micrometers chromium coated rods as authorized contents.

2) Change the brand name from M5 to M5Framatome.

3) Addition of the impact of the variations in the mechanical properties of shells and screws between 40°C and 78°C.

4) Additional justification of the strength of the bolts linked to the half-shells in axial drop.

5) Addition of the analysis of the ageing mechanisms.

6) Revisions to the structural and other analyses based on proposed changes.

4) Update of S9 welds.

The applicant included a brief description of the changes related to this application in document No. NTE-22-006384-000, version 1.0 (DOT, 2022).

The applicant also requested U.S. revalidation of the certificate for the FCC-4 package to the requirements of the International Atomic Energy Agency (IAEA) specific safety requirements No. 6 (SSR-6), Regulations for the Safe Transport of Radioactive Material, 2018 Edition (IAEA, 2018). The staff previously revalidated the certificate for this package per the 2012 edition of the IAEA SSR-6 (IAEA, 2012) (NRC, 2021).

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION The NRC reviewed the information provided to the DOT by Orano in its application for the Model No. FCC-4 package and its supplements against the regulatory requirements of the IAEA SSR-6, 2018 Edition (IAEA, 2018). Based on the statements and representations in the information provided by DOT and the applicant, the staff recommends the revalidation of FCAC No. F/348/AF-96, revision Gw, Model No. FCC-4 package.

1.0 GENERAL INFORMATION REVIEW The staff previously provided a recommendation to DOT on the FCAC No. F/348/AF-96, Model No. FCC-4, package (NRC, 2021), under the IAEA SSR-6 regulations (IAEA, 2018). The staff focused the review of this submittal on the changes made to the design of the Model No. FCC-4 since the revalidation recommendation provided to DOT in 2021 (NRC, 2021) and the FCAC No. F/348/AF-96. The requirements for the contents of a certificate of approval are included in paragraph 838 of the IAEA 838 (IAEA, 2018).

1.1 Package Description 1.1.1 Packaging The Model No. FCC-4 packaging has a length of 5,748 millimeters (mm), an outside diameter of 1,134 mm, and a height of 1,297 mm and a maximum loaded weight of approximately 5,550 kilograms (kg) (DOT, 2022). The packaging is composed of a lower shell (the base of the packaging), an upper shell (packaging cover), a cradle made up of two stringers and connected to the lower shell by rubber shock mountings, and an internal system (frame and doors) leaving space for two cavities to accommodate the authorized and proposed contents. The applicant requested changes as described in the Summary section of the safety evaluation report (SER) and are further evaluated in this SER.

1.1.2 Contents The FCC-4 shipping package is designed to transport unirradiated commercial pressure water reactor fuel assemblies or individual fuel rods. The application defined two versions for the FCC-4 packaging, which vary depending on the length of the fuel rods or full fuel assemblies.

The first design version applies to fuel assemblies composed of a 17x17 square fuel rod array or for non-assembled fuel rods grouped in channels. The second design version applies to fuel assemblies composed of a 16x16 or 18x18 square fuel rod array. The characteristics of the fuel assemblies are defined in the certificate from the French Competent Authority (appendix 1, of the FCAC No. F/348/AF-96, revision Gw).

The contents are described as fuel rod assemblies with sintered pellets in a zirconium alloy cladding that may be pre-oxidized, which can be covered with a layer of chromium up to 30 µm thick, and meets the criteria (e.g., dimensions, weight, material, density, enrichment, and mass ratio) for the FCC-4 packaging. Even though reprocessed uranium is mentioned in the application reprocessed fuel is not allowed by the French certificate and was not evaluated for revalidation by the staff. Moreover, the applicant informed the NRC during a prior revalidation request that there is no intent to transport enriched reprocessed fuel in the FCC-4 packaging (NRC 2021). The staff finds that this is consistent with the contents in annexes 1 through 13 (excluding annex 9, which was not provided) of the approval certificate for the FCC-4 packaging which describes the pellets as enriched natural uranium (ENU). The staff finds the description of the contents of the package in the approval certificate for the FCC-4 packaging to be acceptable 2

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION after reviewing the annexes and the overall approval certificate. Changes related to the authorized contents of the FCC-4 are evaluated in detail in this SER.

1.2 Drawings Section 5.0 of this SER includes the evaluation of the changes to the drawings for the design of the FCC-4 package.

1.3 Evaluation Findings

Based on the information provided in the application and its supplements, the staff finds, with reasonable assurance, that the applicant demonstrated compliance with the requirements in paragraph 838 of the IAEA regulations (IAEA, 2018).

2.0 STRUCTURAL EVALUATION 2.1 General Considerations The purpose of the structural evaluation is to verify that the structural performance of the package meets the regulatory requirements of the IAEA SSR-6 (IAEA, 2018). The applicant performed structural analyses to demonstrate that the strength of the FCC-4 package meets the requirements specified in the IAEA SSR-6 (IAEA, 2018). Specifically, the applicant addressed the FCC-4 package under:

1) Routine conditions of transport (RCT) with package tie-down and lifting,

2) Normal conditions of transport (NCT) under regulatory tests relating to NCT, and

3) Accident conditions of transport (ACT).

A summary of the staffs structural evaluation is provided below.

2.2 Structural Analysis under Routine Conditions of Transport The applicant submitted appendices 2.1-4, FCC 4 - containers for fresh fuel assemblies-Lifting points mechanical verification;1 2.1-5, FCC 4 - containers for fresh fuel assemblies - Stacking behaviour;2 2.1-6, Transportation in FCC container Mechanical aspects related to a change in the fuel assembly materials;3 2.1-11, Justification of the Performance of the FCC-4 Shell Fasteners during a 9-m Vertical Drop (FCC-3/FCC-4);4 and 2.1-14, Variation in the Mechanical Properties of Shell and Screw Steels between -40°C 5) in the application to update chapter 2.1, Structural Analysis, of the safety analysis report (SAR; thereafter, the application). The applicant described the structural analysis for the FCC-4 package lifting under RCT in section 3.2, Package lifting, in chapter 2.1 of the application. Section 3.2 includes the structural analysis for the package lifting to demonstrate the behavior of the FCC-4 package under RCT.

1 Document No. D02-ARV-01-186-616-NPV, revision A (ML22277A753) 2 Document No. DOS-19-021166-013-NPV, version 1.0 (ML22277A754); reference No. D02-ARV-01-186-618-NPV, revision A (ML22277A755) 3 Document No. DOS-19-021165-017-NPV, version 1.0 (ML22271A636) 4 Document No. DOS-19-021165-011-NPV, version 1.0 (ML22271A807) 5 Document No. DOS-19-021165-012-NPV, version 1.0 (ML2271A810) 3 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION The applicant used the SYSTUS finite element (FE) computer program to perform numerical calculations to demonstrate the acceptability of the FCC-4 package design as well as compliance with the applicable regulatory requirements. The applicant assumed the lifting lugs located on the upper shell. Appendix 2.1-4 of the application (document No. DOS-19-021166-012, version 1.0) includes additional assumptions related to the FE analysis (i.e., mesh, material properties, load distribution, and boundary conditions) performed by the applicant. The applicant provided the results of the analysis in a calculation package submitted with the application.6 The applicant compared the calculated stress with the allowable stress for the shell structure, weld, and bolt, and calculated a margin of safety (MS), to find a ratio of (allowable stress/calculated stress - 1.0) (see table 1 below). Since the calculated minimum MS are larger than 0.0, the allowable stress is larger than the calculated stress, which indicates that the FCC-4 package has adequate strength for uplifting under RCT.

Table 1. [Table Withheld per 10 CFR 2.390]

The staff reviewed the results of the structural analysis and finds that the applicant demonstrated that the FCC-4 transport package meets the regulatory requirements of the IAEA SSR-6.

2.3 Structural Analysis Under Normal Conditions of Transport The applicant described the structural analysis for the FCC-4 package stacking test under NCT in subsection 4.2.1, Stacking testing, in chapter 2.1 of the application. Subsection 4.2.1 includes the structural analysis for the stacking test to demonstrate the behavior of the FCC-4 package under NCT.

The applicant used the SYSTUS FE computer program to perform the numerical calculations and demonstrate acceptability of the FCC-4 package design as well as compliance with the applicable regulatory requirements. The applicant provided additional assumptions of the FE analysis such as mesh, material properties, load distribution, and boundary conditions in appendix 2.1-5 of the application. The applicant provided the results of this analysis in a calculation package submitted with the application (see table 2 below).7 Since the calculated MS is larger than 0.0, this indicates that the FCC-4 package has adequate strength for stacking under NCT.

6 Document No. D02-ARV-01-186-617 EN, Rev. A (ML22277A753) 7 Document No. D02-ARV-01-186-618, revision A (ML22277A755) 4 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION Table 2. [Table Withheld per 10 CFR 2.390]

The staff reviewed the results of the structural analysis and finds that the applicant demonstrated that the FCC-4 transport package meets the regulatory requirement of IAEA SSR-6 (IAEA, 2018).

2.4 Structural Analysis Under Accident Conditions of Transport Section 3.3., Accident conditions of transport, in chapter 1.5, Package Performance Characteristics, of the application states that all drop tests were carried out on the full-scale prototype of the package, which is representative of the actual FCC-4 package. The prototype fuel assemblies used during the drop tests were dummy assemblies with cladding made of Zircaloy-4 (Zy-4) (DOT, 2022). The assemblies were filled with materials whose mechanical characteristics were representative of the fuel pellets. Therefore, the structural and geometric characteristics were representative of the actual assemblies transported in the FCC-4 package.

The results of the full-scale prototype drop tests under the ACT were provided in chapter 2.1 of the application, which the staff previously reviewed and validated. The results showed that the enclosure of the prototype package suffered localized deformations but did not impact the containment of the package. Moreover, there was no dispersal of the radioactive contents after completing the regulatory tests for ACT.

In this application, the applicant added additional information of the three appendices (appendices 2.1-6, 2.1-11, and 2.1-14) to chapter 2.1 of the application to demonstrate the adequacy of the FCC-4 package design under ACT by using a computer code with proper inputs (i.e., material properties).

2.4.1 Mechanical Property Changes of Fuel Assembly Materials Appendix 2.1-6 includes the analysis and experimental tests to find the acceptability of the mechanical strength properties of the fuel assembly materials (including fuel claddings made of Zy-4) under ACT over the temperature range from -40°C to 78°C (DOT, 2022). The applicant used the classical elasticity theory and performed experimental tests to analyze the behavior of the materials with considerations of temperature variations and dynamic drop conditions. The calculation package, Transportation in FCC Container Mechanical Aspects Related to a Change in the Fuel Assembly Materials, 8 provided the results of the analysis with the criteria for the acceptable mechanical strength properties (yield strength, ultimate strength, and elongation) in the temperature range between -40°C and 78°C under ACT. The calculation package includes the proposed criteria for the cladding material as shown in table 3.

8 Document No. FFDC04223-NPV EN, revision 4.0 (ML22271A637) 5 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION Table 3. [Table Withheld per 10 CFR 2.390]

Based on the results of the study, the applicant concluded that any material is acceptable for cladding if the material meets one of the five proposed criteria shown in the columns of the table above. However, if a material does not meet one of these criteria, the applicant noted that a dedicated independent study should be undertaken to check the acceptability of the material for cladding.

The staff understood that those five criteria were developed based on the tests on the different specimens (e.g., Zy-4, M5Framatome, Duplex, PCA-2b, and Zr1Nb). However, it was not clear whether the proposed criteria for cladding materials was appliable to other materials, besides the five cladding materials tested and analyzed by the applicant. As a result, the staff requested the applicant to provide the following information (DOT, 2023b):

a) explain whether the proposed five criteria are only applicable to those cladding materials specified in the FCAC, or cladding materials beyond the ones specified in the FCAC; b) technical justifications of the applicability of those proposed criteria to any other materials, if the applicant is proposing the criteria beyond the cladding materials specified in the FCAC to any other materials; and c) the exact information of what types of materials that the applicant is requesting in the application for cladding.

The applicant provided a response to the request for additional information (RAI) stating that the table shown above is a broader list than what is allowed by the FCAC approval. However, the FCAC approval is only limited to Zirconium alloys meeting either of the following design criteria for cladding properties shown in the table below.

Table 4. Material properties design criteria for zirconium alloys.

Criterion Material Properties 1 2 Yield Stress, Rp 0.2% (MPa) 520 250 Ultimate Stress, Rm (MPa) 710 400 At (% on 50 mm) Elongation 12 25 The applicant stated that the properties of Zy-4 cladding material meet criterion 1 and M5Framatome cladding material meets criterion 2 shown in table 4 of this SER. As a result, the applicant is requesting only to use Zy-4 and M5Framatome (changed from M5 to M5Framatome) as cladding materials in this application. The staff reviewed the application and its supplements and finds the material properties of Zy-4 and M5Framatome acceptable. The properties of these materials are based on the tests and analyses, which meet the design criteria for the cladding material properties previously reviewed and recommended for revalidation (NRC, 2021) by the staff with these alloy materials as authorized contents.

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 2.4.2 Mechanical Property Changes of Shell and Screw Steels Appendix 2.1-14 of the application provided the analysis to find changes of the mechanical strength properties for the steels forming the upper shell, the lower shell, and their connecting screws in the FCC-4 package in a temperature range of -40°C to 78°C. The tables in appendix 2.1-14 provided the changes of the mechanical strength properties (i.e., yield and ultimate strength) for the shell and screw materials. Based on the results of their analysis, the applicant concluded that the temperature variation on the mechanical properties of the steels used for the shells and screws has a very minor effect on the energy absorption capacities of the materials (shells and screws) and, therefore, does not significantly impact the drop behavior of the package.

Based on the information provided by the applicant, the staff finds that the mechanical properties of the steels used for shells and screws would not significantly impact the performance of the package during a drop.

2.4.3 Numerical Analysis of the FCC-4 Package for a 9-m Vertical Drop Appendix 2.1-11 of the application provided the analysis to investigate the behavior of the FCC-4 package under ACT using the calculated mechanical strength properties from appendices 2.1-6 and 2.1-14 (DOT, 2022). The applicant used LS-DYNA FE computer program for the structural drop analysis under ACT. The detailed information of the FE analysis (i.e., mesh, element types, material properties, load distribution, and boundary conditions) is provided in appendix 2.1-11 of the application. The applicant performed a 9-m vertical drop analysis was performed using the LS-DYNA FE program for a representation of the prototype 1 test model previously discussed in appendix 2.1-9 of the application.

The results of the FE numerical analysis were provided in tables and figures of appendix 2.1-11 of the application. Based on the results of the FE analysis and a comparison study between the FE numerical analysis and prototype model test, the applicant concluded the following:

a) The two (upper and lower) half-shells of the FCC-4 package remain joined, and b) The results of the numerical analysis using the LS-DYNA FE program verified the results of prototype model:

i) the package suffered localized deformations, but did not impact the safety of the package, and ii) there was no dispersal of the radioactive contents from the FCC-4 transportation package under ACT.

The staff reviewed the numerical drop analysis results, including the supporting structural analyses, and finds that the applicant adequately demonstrated that the FCC-4 package has adequate strength to withstand ACT and meet the regulatory requirements of IAEA SSR-6 (IAEA, 2018).

2.5 Evaluation Findings

Based on the review of the statements and representations contained in the application, the staff concludes that the structural evaluations and the regulatory drop test program have been 7

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION adequately described, and the FCC-4 transport package has adequate structural design to meet the requirements of the IAEA SSR-6 (IAEA, 2018).

3.0 THERMAL EVALUATION The purpose of the thermal evaluation is to assess the packages design against the thermal performance requirements in the IAEA SSR-6 (IAEA, 2018). The applicants proposed changes to the design of the FCC-4 package do not have direct impact on the thermal performance of the package.

3.1 Description of Thermal Design The design of the package is described in the approved FCAC No. F/348/AF-96, revision Gw, dated June 13, 2022, and the application for the Model No. FCC-4, Oranos document No.

DOS-19-021166-000, revision 5.0, dated February 21, 2022.

The applicant did not propose changes to the design features related to the thermal evaluation of the FCC-4 package as part of this request for revalidation.

3.2 Material Properties and Component Specifications Section 5.0 of this SER includes the materials evaluation related to the changes proposed in this action.

3.2.1 Material Properties The applicant did not propose changes to the properties of the materials used to fabricate the FCC-4 packaging as part of this request for revalidation.

3.2.2 Component Specifications The applicant did not propose changes to the specifications of the components for the FCC-4 packaging as part of this application for revalidation.

3.3 General Considerations The application includes descriptions of the analysis models and results of analyses completed to demonstrate the thermal performance of the FCC-4 package. Table 5 of this SER includes the documents containing the analyses related to the thermal performance of the package.

The applicant used the results of these analyses as part of the demonstration that, for the content already authorized by the French competent authority under FCAC No. F/348/AF-96, version Gw, the FCC-4 package meets the thermal requirements in the IAEA SSR-6 (IAEA, 2018).

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION Table 5. Documents related to the thermal performance of the Model No. FCC-4 package.

Document No. Application Title (DOT, 2022) Section DOS-19-021165-013, version 1.0a Chapter 2.2 Thermal Analysis DOS-19021165-018, version 1.0 Appendix Justification by CFD simulation of the thermal (D02-ARV-01-198-455)b 2.2-1 behavior of the FCC container Appendix Additional information for justification of the DOS-13-00081778-202 (REMOVED) 2.2-2 thermal behaviour of the FCC container Behavior of a rod made of M5Framatome or Appendix DOS-18-016472-017, revision 1.0c Zircaloy-4 alloy subjected to the IAEA thermal 2.2-3 test in FCC packaging Appendix Thermal behaviour of the FCC packaging DOS-13-00081778-206, revision 00d 2.2-6 under normal conditions of transport Notes:

a. Document No. DOS-19-021165-013, version 1.0 (ML22271A811)

b. Document No. D02-ARV-01-198-455 (ML22277A770)

c. Document No. DOS-18-016472-017, revision 1.0 (ML22277A771)

d. Document No. DOS-13-00081778-206, revision 00 (ML22271A791) 3.4 Thermal Evaluation under Normal Conditions of Transport 3.4.1 Heat and Cold The changes requested by the applicant and the additional description of ageing considerations added to the safety documentation for the package do not change the thermal performance of the package under the conditions stated for NCT in paragraphs 654, 656, and 657 of the IAEA SSR-6 (IAEA, 2018). Therefore, the package continues to meet these requirements.

3.5 Thermal Evaluation under Accident Conditions of Transport 3.5.1 Initial Conditions Initial conditions for the ACT evaluation are defined in paragraph 728, Thermal test of the IAEA SSR-6 (IAEA, 2018 Edition). The applicant did not propose changes to the initial conditions for the package evaluation as part of this application for revalidation.

3.5.2 Fire Test Conditions The fire test conditions requirements for the evaluation of ACT are defined in paragraph 728(a) of the IAEA SSR-6 (IAEA, 2018). No changes in the fire test conditions for the package evaluation have been observed as part of this request for revalidation.

3.5.3 Maximum Temperatures and Pressure The applicant did not propose changes to the maximum temperatures and pressures for the package as of this request for revalidation.

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.5.4 Maximum Thermal Stresses The applicant did not propose changes to the maximum thermal stresses in the package as part of this request for revalidation.

3.5.5 Analyses Details The applicant did not propose changes to the thermal analyses completed for the FCC-4 package under ACT.

3.6 Evaluation Findings

Based on review of the statements and representations in this request for revalidation, the staff finds that the FCC-4 package conforms to the requirements for Type A fissile packages loaded with fissile materials found in the regulations for the safe transport of radioactive material, IAEA SSR-6 (IAEA, 2018).

4.0 CONTAINMENT EVALUATION The purpose of the containment review is to verify that proposed changes to the FCC-4 package design meet the requirements for evaluating the containment boundary as required in the IAEA SSR-6 (IAEA, 2018). The staff reviewed the application and confirmed that the FCC-4 package containment system had a suitable description for revalidation. The staffs evaluation of the components related to the containment system related to this application is provided below.

4.1 Description of the Containment System The fuel cladding (Zy-4 and M5Framatome cladding) along with the zirconium alloy welded end plugs form the containment boundary for the fissile material in the FCC-4 package. The staff reviewed this information in chapter 1.5, section 2.4, Enclosure, of the application. After reviewing the description, the staff finds the description to be adequate.

The application further describes the fuel rod cladding performance during NCT and ACT, as described below. The staff looked at how the fuel cladding performed under both normal and accident conditions of transport because it is essential to the containment boundary. In chapter 2.1, "Structural Analysis," section 4.3.3, "Representativity of the prototype 2 contents, the applicant noted that the pre-oxidation surface treatment of the cladding does not modify the mechanical characteristics or the structure of the fuel assemblies. The applicant also noted that the fuel cladding may have a chrome coating with a maximum thickness of 30 µm, which is evaluated in section 2.0 of this SER.

Moreover, the applicant noted that there was no dispersion of radioactive contents upon completion of the regulatory tests for NCT in chapter 1.5, section 3.2, Normal conditions of transport, and chapter 2.1, section 4.2.4, Conclusion, of the application. Additionally, in chapter 1.5, section 3.3, "Accident conditions of transport," of the application, the applicant provided results of the satisfactory performance of the Zy-4 for the fuel assembly prototypes during the regulatory drop tests. The results in the prototypes are extended to the rods made of Zirconium and M5Framatome. Furthermore, in chapter 1.5, section 3.3, the applicant noted that the 10 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION numerical analysis of the stresses in the closing screws of the FCC-4 packaging half-shells and prototype 1 justifies the absence of separation of the packaging half-shells by extension for the FCC-4 packaging. The analysis considers the following:

1) accumulation of NCT and ACT regulatory drops,

2) variation of mechanical properties of the materials of the package depending on the temperature, and

3) variations of the tightening torques.

The staff finds the description of the containment boundary for the fissile material to be acceptable.

The applicant summarizes that the fuel cladding was intact, leak tight, and there was no radioactive content dispersion after performing the tests representative of accident drop conditions in chapter 2.1, section 4.3.5, "Drop test results for prototype 2," of the application. In chapter 2.2, Thermal Analysis, section 10, Conclusion, of the application, the applicant reached the conclusion that there was no material dispersion for the Zy-4 and M5Framatome cladding after the tests representative of ACT are completed (DOT, 2022).

The applicant describes the existence of a residual amount of 5 grams (g) of glycerine within the FCC-4 package in appendix 2.2-5, "Analysis of the Impact of Glycerine on the Thermal Safety Analysis," section 5, "Conclusion," of the application. This amount, however, does not bring into question the M5Framatome or Zy-4 fuel rod cladding's mechanical behavior. The staff finds that the fuel cladding will maintain containment under both normal and accident conditions of transport based on the information provided in this section, and it meets the provisions of paragraph 648 of the IAEA SSR-6 (IAEA, 2018).

4.2 General Considerations Section 1 for each of the contents in annexes 1 through 13 of the French certificate for the FCC-4 package design includes a description of the physical state of the contents. The staff notes that there is no annex 9 of the approval certificate for the FCC-4 packaging; therefore, the staff cannot approve an annex 9 of the approval certificate for the FCC-4 packaging. Section 1.1.2 of this SER includes a brief description of the zirconium fuel rods to be transported in the FCC-4 packaging. The staff reviewed the annexes and the overall approval certificate. The staff finds the description of the contents of the package in the approval certificate for the FCC-4 packaging to be acceptable.

4.2.1 Classification as a Type A-Fissile Package In annexes 1 through 13 (excluding appendix 9 which was not provided) of the French Certificate of Approval for the FCC-4 package describes the maximum activity per packaging as less than 1 A2.

The applicant refers to the A2 value of ENU under 20% as unlimited in chapter 1.6, Compliance with regulatory requirements, section 8.1, Article 429, of the application. As a result, the staff concludes that the FCC-4 package is appropriately classified as type A(F) package by the certificate of approval because its contents have less activity than 1 A2.

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION Therefore, the staff finds that the package meets the provisions of paragraph 429 of the IAEA SSR-6 (IAEA, 2018).

Chapter 1.3, Specifications relating to radioactive content, table 1.3-5, Maximum activity of ENU assemblies, shows that the total activity for the types of fuel assemblies and rods is equal to 0 A2. As a result, the staff considers the activity being equal to 0 A2 to be acceptable. The staff concludes that the assemblies and rods can be transported in a type A(F) package because it complies with paragraph 429 of the IAEA SSR-6 (IAEA, 2018).

4.2.2 Gas Generation The applicant states in section 8.11, "Article 644," of chapter 1.6 of the application, that the containment system is made up of the fuel rod cladding and that the rods only contain UO2 pellets and inert gas. According to the applicant, the rods do not contain any liquids or materials which may generate gas by radiolysis or chemical reaction. The staff finds that this description is acceptable as it meets the provisions of paragraph 644 of IAEA SSR-6 (IAEA, 2018).

4.2.3 Radiation Shield In chapter 1.6, in section 8.14, Article 647, the applicant states that the resin found in the support frames and doors makes up most of the radiation shield, which surrounds the containment system. The doors are attached to the support frame and have quick release pins that are resistant to both normal and accident conditions of transport.

The support frame, the doors, and the head and foot plates encircle the cavity. Under any transport circumstances, the fuel assemblies stay in the cavity and consequently inside the radiation shield. The staff review finds that the description of the design of the containment system in reference to the radiation shield meets the provisions of paragraph 647 of the IAEA SSR-6 (IAEA, 2018).

4.3 Containment Evaluation under Accident Conditions of Transport In chapter 1.6, section 12.1.1, Articles 719, 720 and 721: Water spray test, the applicant evaluated the effect of the water spray test and concluded that the test has no impact on the mechanical behavior in the free drop, stacking, or penetration tests. The metallic half-shells and associated gasket prevent water from penetrating inside the packaging during the test.

The conclusions of the test have no impact on the compliance of the package with the regulatory requirements, as a result, staff finds the description to be acceptable.

In chapter 1.6, section 12.1.2, Article 722: Free drop test, the applicant states that after the test in NCT, there is no effect on the compliance of the package with regulatory requirements.

The staff finds the description to be adequate.

In chapter 1.4, "Specification Relating to the Packaging," 9 the applicant noted in section 5.7, "Thermal protection system," that the two half-shells, internal equipment system, and resin present in the doors and frame all function to protect against fire. This resin also enables the 9 Document No. DOS-19-021166-005-NPV, version 1.0 (ML22277A734) 12 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION containment system (i.e., the fuel rod cladding) to be shielded from the effects of the temperatures present during ACT. The staff reviewed the description and found it to be adequate.

4.4 Evaluation Findings

Based on statements and representations in the FCC-4 package application, the staff finds that the containment design has been adequately described and evaluated and that the package design complies with the containment requirements of the IAEA SSR-6 (IAEA, 2018). The staff recommends revalidation of the FCAC No. F/348/AF-96, revision Gw.

5.0 MATERIALS EVALUATION The staff reviewed the information provided in the application and its supplements against the regulatory requirements of the IAEA SSR-6, 2018 Edition (IAEA, 2018). The staff also reviewed the proposed changes to the package as described in the summary section of this SER and section 1.0.

The staff conducted a revalidation review for the FCAC No. F/348/AF-96, revision Gw, Model No. FCC-4, type A(F) package, per the requirements in section VI of the IAEA SSR-6 (IAEA, 2018). The pertinent IAEA SSR-6 requirements10 are listed below:

1) 613. The package shall be capable of withstanding the effects of any acceleration, vibration or vibration resonance that may arise under routine conditions of transport without any deterioration in the effectiveness of the closing devices on the various receptacles or in the integrity of the package as a whole. In particular, nuts, bolts, and other securing devices shall be so designed as to prevent them from becoming loose or being released unintentionally, even after repeated use.

2) 613A. The design of the package shall take into account aging mechanisms.

3) 614. The materials of the packaging and any components or structures shall be physically and chemically compatible with each other and with the radioactive contents. Account shall be taken of their behavior under irradiation.

4) 616. The design of the package shall take into account ambient temperatures and pressures that are likely to be encountered in routine conditions of transport.

5) 639. The design of the package shall take into account temperatures ranging from -40°C to +70 °C for the components of the packaging. Attention shall be given to freezing temperatures for liquids and to the potential degradation of packaging materials within the given temperature range.

6) 640. The design and manufacturing techniques shall be in accordance with national or international standards, or other requirements, acceptable to the competent authority.

10 Requirements listed in 5.0 of this SER correspond to section VI of IAEA SSR-6, 2018 Edition (IAEA, 2018).

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7) 648. A package shall be so designed that if it were subjected to the tests specified in paras 719-724, it would prevent:

a) Loss or dispersal of the radioactive contents; b) More than a 20% increase in the maximum dose rate at any external surface of the package.

8) 673. Fissile material shall be transported so as to:

a) Maintain subcriticality during routine, normal and accident conditions of transport; in particular, the following contingencies shall be considered:

I. Leakage of water into or out of packages, II. Loss of efficiency of built-in neutron absorbers or moderators, III. Rearrangement of the contents either within the package or as a result of loss from the package, IV. Reduction of spaces within or between packages, V. Packages becoming immersed in water or buried in snow; and VI. Temperature changes.

b) Meet the requirements:

I. Of para. 636 except for unpackaged material when specifically allowed by para. 417(e),

II. Prescribed elsewhere in these Regulations that pertain to the radioactive properties of the material, III. Of para. 637 unless the material is excepted by para. 417, and IV. Of paras 676-686, unless the material is excepted by para. 417, 674 or 675.

9) 679. The package shall be designed for an ambient temperature range of -40°C to +38 °C unless the competent authority specifies otherwise in the certificate of approval for the package design.

10) 728. Thermal test: The specimen shall be in thermal equilibrium under conditions of an ambient temperature of 38°C, subject to the solar insolation conditions specified in table 12 and subject to the design maximum rate of internal heat generation within the package from the radioactive contents.

Alternatively, any of these parameters are allowed to have different values prior to, and during, the test, provided due account is taken of them in the subsequent assessment of package response. The thermal test shall then consist of (a) followed by (b).

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION a) Exposure of a specimen for a period of 30 min to a thermal environment that provides a heat flux at least equivalent to that of a hydrocarbon fuel-air fire in sufficiently quiescent ambient conditions to give a minimum average flame emissivity coefficient of 0.9 and an average temperature of at least 800°C, fully engulfing the specimen, with a surface absorptivity coefficient of 0.8 or that value that the package may be demonstrated to possess if exposed to the fire specified.

b) Exposure of the specimen to an ambient temperature of 38°C, subject to the solar insolation conditions specified in table 12 and subject to the design maximum rate of internal heat generation within the package by the radioactive contents for a sufficient period to ensure that temperatures in the specimen are decreasing in all parts of the specimen and/or are approaching initial steady state conditions. Alternatively, any of these parameters are allowed to have different values following cessation of heating, provided due account is taken of them in the subsequent assessment of package response. During and following the test, the specimen shall not be artificially cooled and any combustion of materials of the specimen shall be permitted to proceed naturally.

The staff reviewed the application to determine the adequacy of the materials of construction of the package and the associated technical discussions in justification of compliance with the requirements in the IAEA SSR-6 (IAEA, 2018).

5.1 Drawings The staff reviewed the drawings and chapter 1.4 of the application and verified that the applicant provided an adequate description of the component safety functions, materials of construction, dimensions and tolerances, and fabrication specifications. The staff notes that the low-alloy steels, carbon steels, and stainless steels used in the fabrication of the FCC-4 packaging are specified in chapter 1.4 of the application as conforming to the applicable Association Française de Normalisation (AFNOR) standards.

Based on the evaluation above, the staff finds that the drawings contain sufficient information to describe the design and manufacture the package, and the package meets the requirements in paragraph 640 and 838 (j) and (k) of the IAEA SSR-6 (IAEA, 2018).

5.2 Materials Standards Chapter 1.4 of the application states that the packaging will be fabricated per specification values defined in table 1.4.-1, chapter 1.4 of the application. These values are defined per the latest revision of the cited AFNOR standards. The application also clarified that AFNOR grades defined in table 1.4.-1, chapter 1.4 of the application, may be replaced by grades with at least equivalent mechanical properties (see table 6 below).

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION Table 6. Standards used in the design of steel materials used in the design of the FCC-4.

Packaging Material AFNOR standard(s)

Low-alloy steel and associated grades NF EN 10025 and NF EN 10025-3 Carbon steel and associated grades NF EN A 36-601 Stainless-steel NF EN 10088-3, NF EN 10028-7, and NF A 35-557 The staff confirmed the mechanical properties defined in the application are consistent with those in the AFNOR standards. Based on the evaluation above, the staff finds that the design of the packaging materials are in accordance with applicable standards, and the package meets the requirements in paragraph 640 of IAEA SSR-6 (IAEA, 2018).

Table 7. Standards related to bolting materials used in the design and fabrication of the FCC-4.

Packaging Material AFNOR standard(s)

Low alloy steel (bolting) NF-EN-898.1 and NF-EN-898.2 Bolting materials (general tolerances) NF-E-86-050 and NF-E-02-350 In addition to the standards shown in table 7 of this SER, package fabrication requires minimum guaranteed values for bolt toughness at -40°C, in accordance with AFNOR standard NF EN 10113-1, and consistent with the requirements in paragraphs 639 and 679 of IAEA SSR-6 (IAEA, 2018).

Appendix 1.7 of the application includes specific torque values for the bolts to be used to secure the two half-shells, including specific torquing methods to distribute the preloads in a homogenous way. The staff reviewed closure procedures and finds this meets the requirements of paragraph 613.

Chapter 1.5, section 2.7; chapter 1.6, section 6.8; and appendix 1.4-5, Analysis of the Aging Mechanisms of the FCC3 and FCC4 Package Models, of the application describes how the package meets the requirements of paragraph 613A of IAEA SSR-6 (IAEA, 2018) related to aging mechanisms. The appendix addresses the effects of the following on the package:

1) moisture (external environment),

2) temperature on the materials,

3) irradiation due to the presence of radioactive material, and

4) fatigue caused by the transport method.

The analysis concluded that the safety functions of the FCC-3 and FCC-4 are not deteriorated due to aging over a minimum period of 37 years (defined in the fatigue analysis). The staff reviewed the analysis of the aging mechanisms and confirmed that the package meets the requirements of 613A of the IAEA SSR-6 (IAEA, 2018).

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.3 Weld Design and Inspection Table 1.4-5, chapter 1.4, of the application lists the packaging welds defined as important for safety (location, type of weld, type of inspection). The applicant noted that weld inspection activities include visual inspection and dye penetrant testing, which is to be completed per CODAP (Code Français de Construction des Appareils Pression; French construction Code for pressurized vessels), section I or another equivalent code. The applicant provided English translations of the Translation of CODAP 2005 Code defining the acceptance criteria for visual inspection (Division 1, part I, Controls and Inspection, appendix I1.A1) and dye penetrant (appendix I1. A2). Chapter 1.7 of the application requires that these inspection activities be performed by qualified personnel, consistent with the applicants Quality Assurance (QA) program. The staff considers the methods and acceptance criteria for packaging welds to be acceptable.

Based on the evaluation above, the staff finds that the manufacture of the package is in accordance with applicable standards, and the package meets the requirements in paragraph 640 of the IAEA SSR-6 (IAEA, 2018).

5.4 Mechanical Properties 5.4.1 Low-Alloy and Carbon Steels Chapter 1.4 of the application includes a description of the mechanical properties of low-alloy steel materials (including bolting), and carbon steel used in the packaging will be fabricated per the specification values in the latest revision of the AFNOR standards listed in table 1.4-1 of chapter 1.4 of the application. The staff confirmed the temperature-dependent mechanical properties of the material grades in the pertinent AFNOR standards with those defined in chapter 1.4 of the application.

Appendix 1.4-3, Summary of Impact Strength Tests, of the application documents the toughness tests that were carried out on the low-alloy steel material used in the package shells (DOT, 2022). The staff confirmed that the test results demonstrate that the minimum toughness values conform with those defined in chapter 1.4 of the application and the referenced AFNOR standard. Chapter 1.4 of the application also describes that the procurement requirements for low-alloy steel bolting will provide for minimum guaranteed toughness values at -40°C (-40°F) by means of specific product tests, in accordance with the applicable AFNOR standard. The staff confirmed that the minimum guaranteed toughness values defined in chapter 1.4 are consistent with the appropriate AFNOR standard.

Appendix 2.1-14 of the application documents the evaluation of the variation in the mechanical properties of shell and screw steels between -40°C and 78°C (DOT, 2022). The applicant found that the mechanical properties of shell and screw steels, between the temperatures -40°C and 78°C, had a very slight effect on the energy absorption capacities of the materials and did not significantly impact the drop behavior of the packaging. Appendix 2.1-11 also evaluated the variation in the mechanical properties versus temperature with respect to the vertical drop behavior of the packaging in the analyzed configuration. The staff reviewed the evaluation in appendix 2.1-14 regarding the effect of temperature variation on energy absorption and appendix 2.1-11 on vertical drop behavior and confirmed that energy absorption is not significantly impacted and that the package meets the requirements in paragraphs 639, 640, and 679 of the IAEA SSR-6 (IAEA, 2018).

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.4.2 Stainless Steels Chapter 1.4 of the application states that the mechanical properties of stainless-steel materials used in the packaging will be fabricated per the specification values in the latest revision of the AFNOR standards listed in table 1.4-1 of chapter 1.4 of the application. The staff confirmed the temperature-dependent mechanical properties of the material grades in the pertinent AFNOR standards with those defined in chapter 1.4 of the application.

The applicant noted that the stainless-steel materials used in the package design do not exhibit brittle fracture at low service temperatures consistent with the requirements in paragraphs 639, 640, and 679 of the IAEA SSR-6 (IAEA, 2018). The staff finds the applicants assessment acceptable.

5.4.3 Wood (Impact Limiter)

In chapter 1.4 of the application, the applicant notes that the axial shock absorbers are fabricated of a stainless-steel enclosure containing balsa wood. Table 1.4-1 of the application specifies the density, moisture content, and crush strength of the impact limiter wood material.

Table 1.4-1 of the application also defines that the wood grain direction must adhere to the specification in the design drawing. The staff reviewed these material properties used in the applicants mechanical calculations and confirmed that the properties are either conservative or consistent with those in the technical literature.

Based on the evaluations above, the staff finds that the mechanical properties of materials used in the structural analysis are consistent with applicable standards and values in the technical literature, and the package meets the requirements in paragraphs 616, 639, 640, 648, and 679 of the IAEA SSR-6 (IAEA, 2018).

5.4.4 Evaluation Findings Based on the evaluations included in the section 5.4 of this SER, the staff finds that the properties of materials provided by the applicant are acceptable, and the package meets the requirements in paragraphs 639 and 640 of the IAEA SSR-6 (IAEA, 2018).

5.5 Thermal Properties of Materials The staff reviewed the thermal calculations in chapter 2.2, Thermal Analysis, of the application and verified that the thermal properties of the materials are consistent with values available in the technical literature.

5.6 Criticality Control Materials The doors and frame of the packaging encase a neutron-absorbing polymer resin used to maintain subcriticality. Chapter 1.4 of the application states that a qualified process (i.e., controlled injection procedure) is used to fill the packaging cavities with the resin.

Table 1.4-1 of chapter 1.4 of the application defined the properties of the resin material used in the packaging. More specifically, the applicant defined property specifications for composition (hydrogen and boron content), density, thermal conductivity, and heat capacity, which are not specific to a French standard.

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION In chapter 2.5 of the application, the applicant justified the suitability of the resin specifications provided in table 1.4-1 of chapter 1.4 of the application to demonstrate compliance with the thermal test per paragraph 728 of the IAEA SSR-6 (IAEA, 2018). More specifically, appendix 2.2-4 of the application described test results to characterize the loss of hydrogen and boron under conservative fire test conditions. The tests demonstrated that the resin remained largely intact when exposed to a direct flame, forming a surface char layer that protects the material beneath. Further, the tests demonstrate that the resin retains adequate thickness to ensure the sub-criticality of the package array during a fire accident. The applicant used the test data to adjust the neutron absorption performance of the resin in the criticality analysis to account for potential degradation during a fire accident.

The staff reviewed the thermal test results and verified that that the applicant applied appropriate penalties to the neutron absorption performance of the resin in the criticality analysis to account for material changes in a fire accident. The staff also concluded that the resin is expected to remain intact and not relocate under ACT.

Based on the evaluations above, the staff finds that the properties of the neutron absorbing materials used in the criticality analysis are acceptable, and the package meets the requirements in paragraphs 673 and 728 of the IAEA SSR-6 (IAEA, 2018).

5.7 Corrosion and Chemical Reactions In chapter 1.6 of the application, the applicant noted that the constituent materials of the packaging and all internal components or structures were chosen to be physically and chemically compatible with each other and with the intended radioactive contents of the package. The staff reviewed the packaging materials and service environments to verify that adverse reactions will not prevent the package from performing its safety functions.

Also, in chapter 1.6 of the application, the applicant noted that the internal equipment of the package is made exclusively of stainless steel, which protects these parts against corrosion risks. The staff reviewed the materials of the packaging and concluded that the stainless steel is compatible with the air environments to which the surfaces of the packaging are exposed.

Based on this evaluation, the staff finds that the applicant adequately considered the corrosion resistance of the internal equipment.

For low-alloy and carbon steel components, table 1.4-1 of the application indicates that a "corrosion inhibitor + paint" coating is applied to the entirety of the upper and lower shells (inside and out). With respect to these components, chapter 1.7 of the application specifies that, before each shipment, there is a provision to check for the absence of flaking paintwork on the uninterrupted sections of the packaging to detect and remedy incipient corrosion. Further, chapter 1.8 of the application details a periodic maintenance program, which specifies that the FCC-4 packaging is required to undergo an inspection and maintenance operation that includes a check for paint defects and rework deficient areas on the internal and external surfaces of the shells if missing paintwork is identified.

Regarding the aluminum packing shims, in appendix 1.3-1 of the application, the applicant notes that shims are used when the FCC-4 packaging transports unassembled fuel rods in the accompanying fuel rod boxes. The staff reviewed the design drawings in appendix 1.3-2 of the application and applicable sections of the application to evaluate the effects, if any, of intimate contact between aluminum packing shims and the materials in the FCC-4 package. The staff finds that galvanic corrosion between the aluminum packing shims and the stainless steel 19 OFFICIAL USE ONLY - PROPRIETARY INFORMATION

OFFICIAL USE ONLY - PROPRIETARY INFORMATION components of the FCC-4 package is not expected because water is effectively sealed off under NCT. Also, visual inspections are to be performed of the payload cavity prior to loading and following off-loading, which provide reasonable assurance that any corrosion will be detected in a timely manner.

Regarding potential radiolytic gas generation, the staff reviewed the content of the fuel rods and the environmental conditions. The authorized content for the FCC-4 is type A(F), including 232U, 234U, 235U, 236U, which are dominantly alpha emitters, and gadolinium oxide (Gd O ). Some 2 3 daughter products decay via beta emission, but they do not emit significant gamma rays. Inside unirradiated fuel rods, there is no water and, therefore, there is no radiolytic gas generation.

Very little/low strength gamma may be present outside cladding where very small amounts of water are present. Therefore, the staff finds that gas generation from radiolysis is not an issue with unirradiated fuel.

Based on the evaluations above, the staff finds that the package design, inspections, and maintenance activities adequately prevent adverse reactions that may affect the ability of the package to perform its safety functions, and the package meets the requirements in paragraphs 614 and 644 of IAEA SSR-6 (IAEA, 2018).

5.8 Content Integrity: Fresh (Unirradiated) Fuel Cladding In chapter 1.5 of the application, the applicant noted that drop tests were carried out on full-scale prototypes with dummy fuel assemblies with rods made of Zy-4 and filled with a material which had mechanical characteristics representative of the fuel pellets. The applicant clarified that the structural and geometric characteristics of the dummy assemblies were identical to those of the production assemblies in the allowable contents of the FCC-4 package.

Chapter 2.1 of the application states that rods cladded with Zy-4 may also contain a maximum thickness of 30µm of Cr. The coating is applied using physical vapor deposition (PVD)

Magnetron sputtering process with argon plasma. The applicant stated that hydrogen content would be below 50 parts per million (ppm) based on the previous use of the process.

Appendix 2.1-6 of the application provided an evaluation of the mechanical properties of Zy-4 and M5Framatome (with or without chrome coating) clad fuel rods from -40°C to 78°C at strain rates representative of the dynamic loadings during drops. The applicant also provided an evaluation of the bending strength of fuel rods cladded with Zy-4 and M5Framatome (with or without chrome coating). A separate evaluation was performed on the 15µm of Cr coated M5Framatome and 30µm of Cr coated M5Framatome to determine the impact of the coating on the bending strength justification of the cladding for the flat drop from 9 meters (m). The staff reviewed the evaluation and found for 15µm, the average mechanical properties are at least as high as uncoated M5Framatome, and for 30µm, the average and minimal mechanical properties are at least as high as uncoated M5Framatome, and therefore there is no unfavorable impact. The applicant conducted bending tests on Zy-4 and M5Framatome -clad fuel rods. The results of the deformation capacity of the M5Framatome rods were compared to and exceeded the calculated bending loads during a 9-m drop, which was used to define the minimum acceptable mechanical properties for M5Framatome cladding (with or without chrome coating) to ensure containment is maintained during a drop accident. The staff reviewed the process described for the chrome coating and found that as the hydrogen uptake would be below 50 ppm, the base cladding material and mechanical properties are not significantly affected by the coating process. The staff reviewed the aforementioned evaluations and test results and found the fuel cladding acceptable in meeting paragraph 673 of IAEA SSR-6 (IAEA, 2018).

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OFFICIAL USE ONLY - PROPRIETARY INFORMATION In appendix 2.2-2 of the application, the applicant evaluated the behavior of the package contents during the thermal test (fire accident) scenario, which demonstrated that the fuel cladding retained its integrity. Appendix 2.2-3 of the application also includes an evaluation of the thermal-mechanical behavior of the fuel rods during the thermal test, which accounted for the maximum temperatures of the hottest rods for the M5Framatome (with or without chrome coating) and Zy-4 claddings as per appendix 2.2-2, the azimuthal thermal gradient, and the presence of pellets inside the rods. The applicant demonstrated that the risk of ballooning and bursting of the cladding under the effect of creep can be ruled out for both cladding alloys.

Based on the staffs review of the mechanical and thermal tests of the packaging and fuel contents, the staff finds that the fuel cladding is capable of maintaining the fuel in its analyzed configuration during normal and accident conditions of transport, and the package meets the requirements in paragraphs 673 and 728 of IAEA SSR-6 (IAEA, 2018).

5.9 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 FCC-4 package and that the package meets the requirements of the IAEA SSR-6 (IAEA, 2018).

6.0 CRITICALITY SAFETY EVALUATION The purpose of the criticality safety evaluation is to verify that the FCC-4 design, and proposed changes meet the regulatory requirements of IAEA SSR-6 (IAEA, 2018). The staffs basis for revalidation is based on the package meeting the paragraphs that apply to criticality safety regulations within these regulations paragraphs 501(c), 526, 673, 682, 684, 685, 686, 716, 814, 815, and 816 of the IAEA SSR-6 (IAEA, 2018). The package was not evaluated for air transport therefore, the package is not authorized for air transport.

6.1 Design Description The applicant requested U.S. revalidation of the certificate for the Model No. FCC-4 package to the requirements of the IAEA, SSR-6 (IAEA, 2018). The staff previously revalidated the certificate for this package in 2021 (NRC, 2021). The only significant changes to the package design with respect to criticality safety are the following:

1) the removal of the option to transport loose rods without axial and radial support, and

2) the addition of chromium-coated cladding.

Removal of an allowed content configuration, specifically loose rods without axial and radial support, has no effect on the criticality safety of the package. The addition of a chromium coating on each fuel rod in the assembly effectively increases the cladding outer diameter, which decreases the amount of water that can be present in the assembly. Since light water reactor fuel assemblies are under-moderated, reducing water in the assembly will decrease reactivity of the fuel. Therefore, the criticality safety analysis for the previous revalidation recommendation (NRC, 2021) remains applicable and conservative for this content change.

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6.2 Evaluation Findings

The staff reviewed the certificate for the Model No. FCC-4 package, as well as the applicants initial assumptions, model configurations, analyses, and results in the application, including the application. The staff finds that the changes requested by the applicant will not affect the criticality safety of the package. Therefore, the staff finds with reasonable assurance that the package, with the requested contents, will meet the criticality safety requirements of paragraphs 501(c), 526, 673, 682, 684, 685, 686, 716, 814, 815, and 816 of the IAEA SSR-6 (IAEA, 2018).

7.0 QUALITY ASSURANCE The purpose of the QA review is to verify that the proposed changes to the package design meet the requirements of the IAEA SSR-6 (IAEA, 2018). The staff reviewed the description of the QA program for the Model No. FCC-4 package against the standards in the IAEA SSR-6 (IAEA, 2018).

7.1 Evaluation of the Quality Assurance Program The applicant developed and described a QA program for activities associated with transportation packaging components that are important to safety. Those activities include design, procurement, fabrication, assembly, testing, modification, maintenance, repair, and use.

The applicants description of the QA program (i.e., management system and compliance assurance programs in the IAEA SSR-6, 2018 Edition (IAEA, 2018)) meets the requirements of the applicable IAEA SSR-6 (IAEA, 2018). The staff finds the QA program description acceptable, since it allows implementation of the associated QA program for the design, procurement, fabrication, assembly, testing, modification, maintenance, repair, and use of the Model No. FCC-4 transportation package.

The staff finds, with reasonable assurance, that the QA program for the FCC-4 transportation packaging meets the requirements in the IAEA SSR-6 (IAEA, 2018) by encompassing the following:

1) design controls,

2) materials and services procurement controls,

3) records and document controls,

4) fabrication controls,

5) nonconformance and corrective actions controls,

6) an audit program, and

7) operations or programs controls, as appropriate.

The staff finds with reasonable assurance that these controls are adequate to ensure that the package will allow safe transport of the radioactive material authorized in this approval.

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7.2 Evaluation Findings

Based on the review of the statements and representations in the Model No. FCC-4 package application and as discussed in this SER section, the staff has reasonable assurance that the FCC-4 package meets the requirements in the IAEA SSR-6 (IAEA, 2018). The staff recommends the revalidation of FCAC F/348/AF-96, revision Gw.

8.0 MAINTENANCE PROGRAM The evaluation of maintenance activities is included in sections 2.0, 4.0, and 5.0 of this SER.

CONDITIONS The staff recommends the revalidation of FCAC No. F/348/AF-96, revision Gw, for the Model No. FCC-4 package, with the following additional condition:

Transport by air is not allowed.

CONCLUSION Based on the statements and representations contained in the documents referenced above, and the condition listed above, the staff concludes that the changes to the Model No. FCC-4 package, FCAC No. F/348/AF-96 package, meet the requirements of IAEA SSR-6, 2018 Edition (IAEA, 2018).

Issued with letter to R. Boyle, U. S. Department of Transportation, on August 8, 2023.

REFERENCES (IAEA, 2012) International Atomic Energy Agency (IAEA), SSR-6, Regulations for the Safe Transport of Radioactive Material, Revision 1, 2012 Edition, ML22083A077.

(IAEA, 2018) International Atomic Energy Agency (IAEA), SSR-6, Regulations for the Safe Transport of Radioactive Material, Revision 1, 2018 Edition.

(NRC, 2021) Letter from John B. McKirgan, U.S. Nuclear Regulatory Commission (NRC) to Richard W. Boyle, U.S. Department of Transportation (DOT), April 22, 2021, Agencywide Documents Access and Management System (ADAMS)

Accession No. ML2103A434, ML21103A431.

(DOT, 2022) Letter from Richard W. Boyle, DOT, to Director, Division of Fuel Management, NRC, August 18, 2022, ML22277A733, ML22277A716.

(DOT, 2023a) Letter from Richard W. Boyle, DOT, to Norma Garcia Santos, NRC, January 26, 2023, ML23026A302, ML23025A208.

(DOT, 2023b) Letter from Richard W. Boyle, DOT, to Norma Garcia Santos, NRC, May 31, 2023, ML23205A093, ML23164A174.

23 OFFICIAL USE ONLY - PROPRIETARY INFORMATION