Text

Enclosure 2 - Safety Evaluation Report for Revision No. 0 of the Certificate of Compliance No. 9382
	ML23275A040
Person / Time
Site:	07109382
Issue date:	10/23/2023
From:	; Storage and Transportation Licensing Branch
To:	; TN Americas LLC
Shared Package
ML23275A038	List: ML23275A039; ML23275A040; ML23275A041;
References
	EPID L-2021-NEW-0000
	Download: ML23275A040 (81)
	v • d • e

Enclosure 2 SAFETY EVALUATION REPORT Docket No. 71-9382 Model No. TN-EAGLE Package Certificate of Compliance No. 9382 Revision No. 0

SUMMARY

By letter dated December 30, 2020 (Agencywide Documents Access and Management System

[ADAMS] Accession No. ML20365A018), as supplemented on April 29, 2021 (ML21119A307),

TN Americas LLC (TN or the applicant) submitted an application for a certificate of compliance (CoC) for the Model Nos. TN Eagle SC and TN Eagle LC spent fuel packages. On December 9, 2022, TN responded (ML22343A152) to U.S. Nuclear Regulatory Commission (NRC) staffs request for additional information dated May 13, 2022 (ML22112A058). TN provided a revised and updated application on September 20, 2023 (ML23263B108).

The Model No. TN Eagle packaging consists of a forged body, with a bottom closure system including the bottom plate, and sealed at the top with the primary lid. The neutron shielding resin blocks are placed in the lodgments of the shielding rings which are designed to shrink fit tightly against the steel shell surfaces, thus improving the heat transfer across the neutron shield. A personnel barrier prevents access to the outer surfaces of the cask body.

The containment boundary of the TN Eagle package consists of the forged body, a primary lid, a lid port plug, a ram access cover plate, and the associated closure seals. The containment, assembled with bolted connections with no associated welds, maintains an inert atmosphere (helium) in the cask cavity, and is designed to meet the leak-tight criteria per American National Standards Institute (ANSI) N14.5-2014. The lid port plug seal, and the inner seals in the primary lid and ram access cover plate, are the primary containment boundary seals. Outer seals are provided in the primary lid and ram access cover plate to facilitate leak testing of the inner containment seals.

Gamma shielding is performed by the forged cask body and shielding ring steel and, at the cask ends, by the steel top and bottom assemblies of the cask and axial ends of the dry shielded canisters (DSCs). Neutron shielding is performed by the borated VYAL-B resin blocks in the shielding rings surrounding the packaging body, by VYAL-B resin plates at both ends.

Criticality safety is maintained by the fixed geometry of the basket, as well as by the borated aluminum, aluminum/B4C metal matrix composite, or Boral neutron poison plates present between each spent fuel assembly location.

The TN Eagle package is authorized to transport the EOS 89BTH, EOS 37PTH, FO-FC-FF, 24PT1, 32PTH1, 24PT4, and 32PT DSCs. The FO-FC-FF and 24PT1 DSCs were previously evaluated and approved for transportation in the TN NUHOMS MP-187 Multi-Purpose Cask (Certificate of Compliance No. USA/9255/B(U)F-85). The 32PTH1, 24PT4, and 32 PT DSCs were previously evaluated and approved for transportation in the TN NUHOMS-MP197HB package (Certificate of Compliance No. USA/9302/B(U)F-96).

2 The TN Eagle package has two configurations: TN Eagle LC (large canister) and TN Eagle SC (standard canister).

Staff requested the applicant to include the transportation frame in the numerical model of the TN Eagle package for drop analyses conducted under hypothetical accident conditions (HAC) to obtain assurance that the response of the impact limiters remains uninfluenced by the presence of the transportation frame. The current analysis did not include the potential influence of the frame on the response of the package: as such, the frame may impact the rigid surface prior to the engagement of the impact and if the frame is considered, the deformation of the impact limiters may be constrained. The applicant did not respond at this time and the certificate of compliance includes a condition to address that concern.

NRC staff reviewed the application, as supplemented, using the guidance in NUREG-2216. The package was evaluated against the regulatory standards in Title 10 of the Code of Federal Regulations (10 CFR) Part 71, including the general standards for all packages and the performance standards specific to fissile material packages under normal conditions of transport and hypothetical accident conditions. The analyses performed by the applicant demonstrate that the package provides adequate structural and thermal protection to meet the containment, shielding, and criticality requirements after being subject to the tests for normal conditions of transport and hypothetical accident conditions.

Based on the statements and representations in the application, and the conditions listed in the CoC, the staff concludes that the package meets the requirements of 10 CFR Part 71.

EVALUATION 1.0 GENERAL INFORMATION The Model No. TN Eagle design consists of two configurations, i.e., the TN Eagle LC and TN Eagle SC, with both having the same overall dimensions. However, because of different outside diameters for the DSCs, an inner sleeve is used for smaller diameter DSCs for the TN Eagle SC. The thickness of the forged cask body is different for the TN Eagle SC and TN Eagle LC to allow the loading of larger diameter DSCs.

For both models, the packaging overall length is 8100 mm (318.90 in.) with the impact limiters and 5598 mm (220.39 in.) without the impact limiters, and the impact limiter outside diameter is 3550 mm (139.76 in.) while the forged cask body outside diameter is 2142 mm (84.33 in.).

The TN Eagle SC has (i) a forged cask body inner diameter of 1850 mm (72.83 in.), (ii) a cavity length of 5061 mm (199.25 in.), and (iii) and empty weight of 90,000 kg (198,416 lb). The TN Eagle LC has (i) a forged cask body inner diameter of 1940 mm (76.38 in.), (ii) a cavity length of 5086 mm (200.25 in.), (iii) and an empty weight of 87,400 kg (192,684 lb).

Each DSC type consists of a stainless-steel cylinder with a double seal-welded closure, and with similar basket structural materials and geometry, as described in the safety analysis report (SAR). Criticality safety is maintained by the fixed geometry of the basket, as well as by the borated aluminum, aluminum/B4C metal matrix composite, or Boral neutron poison plates present between each spent fuel assembly location.

3 The TN Eagle package is designed to transport DSCs that were designed, fabricated, loaded, and maintained under storage licenses. The applicant requested the TN Eagle package be authorized to transport the EOS 89BTH, EOS 37PTH, FO-FC-FF, 24PT1, 32PTH1, 24PT4, and 32PT DSCs. The FO-FC-FF and 24PT1 DSCs were previously evaluated and approved for transportation in the TN NUHOMS MP-187 Multi-Purpose Cask (Certificate of Compliance No.

USA/9255/B(U)F-85). The 32PTH1, 24PT4, and 32 PT DSCs were previously evaluated and approved for transportation in the TN NUHOMS-MP197HB package (Certificate of Compliance No. USA/9302/B(U)F-96).

The EOS-37PTH and EOS-89BTH DSCs will be transported in the TN Eagle LC, and the 32PT, 32PTH1, 24PT1, 24PT4, and FO/FC/FF DSCs will be transported in the TN Eagle SC. Several types of boiling water reactor (BWR) fuel assemblies (FAs) with or without fuel channels or pressurized water reactor (PWR) FAs with or without control components (CCs) can be contained in the DSCs.

The TN Eagle package is designed to transport a maximum of 89 BWR and 37 PWR fuel assemblies in DSCs. Specific DSC contents are summarized below:

EOS 89BTH - designed to transport up to 89 intact or reconstituted 7x7, 8x8, 9x9, or 10x10 BWR spent fuel assemblies with or without channels. Fuel assembly specifications are contained in table 1.6.2-2. Fuel assembly maximum lattice average initial enrichment and neutron poison requirements are given in table 1.6.2-3.

EOS 37PTH - designed to transport up to 37 intact or reconstituted PWR spent fuel assemblies, with or without control components, up to 8 damaged PWR fuel assemblies with the remaining fuel intact, or up to 4 failed PWR fuel assemblies with the remaining fuel intact. Fuel assembly specifications are given in table 1.6.1-15. Initial enrichment, burnup, and cooling time requirements determined in the burnup credit analysis are given in the loading curve for intact and damaged fuel configurations in tables 1.6.1-2 through 1.6.1-13.

NUHOMS-24PT4 - designed to transport up to 24 intact or reconstituted Combustion Engineering 16x16 PWR spent fuel assemblies, or up to 12 damaged assemblies in failed fuel cans, with the remaining fuel intact. Fuel assembly design specifications are contained in table 1.6.3-3. Fuel assembly maximum initial enrichment and neutron poison requirements are given in table 1.6.3-4.

NUHOMS-32PT - designed to transport up to 32 intact or reconstituted PWR spent fuel assemblies, with or without control components. Fuel assembly specifications are given in table 1.6.4-3. Initial enrichment, burnup, cooling time and poison rod assembly requirements determined in the burnup credit analysis are given for all fuel assembly classes except CE14x14 in the loading curve in table 1.6.4-6. Initial enrichment, burnup, and cooling time requirements determined in the burnup credit analysis are given for intact CE14x14 class fuel assemblies in table 1.6.4-7, for damaged CE14x14 class fuel assemblies in table 1.6.4-8, and for failed CE14x14 class fuel assemblies in table 1.6.4-9. Poison rod assembly requirements for loading are given in table 1.6.4-5.

NUHOMS-32PTH1 - designed to transport up to 32 intact or reconstituted PWR spent fuel assemblies, with or without control components, or up to 16 damaged fuel assemblies with the remaining fuel intact. Fuel assembly specifications are given in table 1.6.5-4. Initial enrichment, burnup, cooling time, and neutron absorber type requirements determined in

4 the burnup credit analysis are given in the loading curves for intact and damaged fuel configurations in tables 1.6.5-7 and 1.6.5-8, respectively.

NUHOMS-FO-FC-FF DSCs - designed to transport up to 24 intact or reconstituted B&W 15x15 PWR spent fuel assemblies in the FO or FC DSCs, or up to 13 damaged or failed Babcock & Wilcox (B&W) 15x15 fuel assemblies in failed fuel cans in the FF DSC. The maximum enrichment for fuel transported in these DSCs is 3.43 weight percent U-235. Fuel assembly design specifications are contained in table 1.6.6-2.

NUHOMS-24PT1 - designed to transport up to 24 intact WE 14x14 stainless steel clad UO2 PWR spent fuel assemblies, or up to 24 intact WE 14x14 zirconium alloy clad mixed oxide (MOX) damaged or failed fuel assemblies. This DSC may also transport up to 4 damaged or failed UO2 fuel assemblies, or 1 MOX fuel assembly, in failed fuel cans, with the remaining fuel intact. Fuel assembly specifications, including maximum initial enrichment for UO2 fuel assemblies and uranium and plutonium contents for MOX fuel assemblies, are given in table 1.6.7-1. Fuel assembly characteristics important for criticality safety are given in table 1.6.7-2.

Damaged fuel assemblies must be placed in basket locations with top and bottom end caps in DSC types for which damaged fuel is authorized. The applicant modeled damaged fuel in various configurations to find the most reactive condition. The applicant properly addressed criticality safety of arrays of spent fuel transportation packages and the package has a criticality safety index (CSI) of 0.0.

The package is constructed and assembled in accordance with the following TN Americas LLC, Drawing numbers:

TN Eagle Package Drawings TN EAGLE01-1100 Rev 0 TN Eagle LC (Large Canister) and TN Eagle SC (Standard Canister) Transport Package (7 Sheets)

EOS Drawings EOS01-71-1000 Rev 0 NUHOMS EOS System Transportable Canister 37PTH DSC Main Assembly (7 sheets)

EOS01-71-1001 Rev 0 NUHOMS EOS System Transportable Canister 37PTH DSC Shell Assembly (2 sheets)

EOS01-71-1005 Rev 0 NUHOMS EOS System Transportable Canister 89BTH DSC Main Assembly (7 sheets)

EOS01-71-1006 Rev 0 NUHOMS EOS System Transportable Canister 89BTH DSC Shell Assembly (2 sheets)

EOS01-71-1010 Rev 0 NUHOMS EOS System Transportable Canister 37PTH DSC Basket Assembly (15 sheets)

5 EOS01-71-1011 Rev 0 NUHOMS EOS System Transportable Canister 37PTH Basket Transition Rails (6 sheets)

EOS01-71-1020 Rev 0 NUHOMS EOS System Transportable Canister 89BTH DSC Basket Assembly (9 sheets)

EOS01-71-1021 Rev 0 NUHOMS EOS System Transportable Canister 89BTH Basket Transition Rails (7 sheets) 24PT1 Drawings NUH24PT1-71-1000 Rev 0 NUHOMS 24PT1-DSC Main Assembly (6 sheets) 24PT4 Drawings NUH24PT4-71-1001 Rev 1 NUHOMS 24PT4 Transportable Canister for PWR Fuel Basket Assembly (5 sheets)

NUH24PT4-71-1002 Rev 1 NUHOMS 24PT4 Transportable Canister for PWR Fuel Main Assembly (8 sheets)

NUH24PT4-71-1003 Rev 0 NUHOMS 24PT4 Transportable Canister for PWR Fuel Failed Fuel Can (4 sheets) 32PT Drawings NUH32PT-71-1000 Rev 1 NUHOMS 32PT Transportable Canister for PWR Fuel Summary Dimensions (1 sheet)

NUH32PT-71-1001 Rev 2 NUHOMS 32PT Transportable Canister for PWR Fuel Main Assembly (5 sheets)

NUH32PT-71-1002 Rev 2 NUHOMS 32PT Transportable Canister for PWR Fuel Shell Assembly (3 sheets)

NUH32PT-71-1003 Rev 2 NUHOMS 32PT Transportable Canister for PWR Fuel A Basket Assembly (16 Poison/16 Compartment Plates) (8 sheets)

NUH32PT-71-1004 Rev 2 NUHOMS 32PT Transportable Canister for PWR Fuel Aluminum Transition Rail - R90 (2 sheets)

NUH32PT-71-1005 Rev 2 NUHOMS 32PT Transportable Canister for PWR Fuel Aluminum Transition Rail -R45 (1 sheet)

6 NUH32PT-71-1006 Rev 2 NUHOMS 32PT Transportable Canister for PWR Fuel A/B/C/D Basket Assembly (20 Poison/12 Compartment Plates) (6 sheets)

NUH32PT-71-1007 Rev 2 NUHOMS 32PT Transportable Canister for PWR Fuel A/B/C/D Basket Assembly (24 Poison/8 Compartment Plates) (8 sheets)

FO and FC Drawings DWG-NUH24P-FOFC-71-1000 Rev 0 NUHOMS FO-DSC & FC-DSC for PWR Fuel Main Assembly (5 sheets)

FF Drawings DWG-NUH24P-FF-71-1000 Rev 0 NUHOMS FF-DSC for PWR Fuel Main Assembly (5 sheets) 32PTH1 Drawings NUH32PTH1-71-1000 Rev 2 NUHOMS 32PTH1 Transportable Canister for PWR Fuel Main Assembly (4 sheets)

NUH32PTH1-71-1001 Rev 2 NUHOMS 32PTH1 Transportable Canister for PWR Fuel Basket Shell Assembly (5 sheets)

NUH32PTH1-71-1002 Rev 2 NUHOMS 32PTH1 Transportable Canister for PWR Fuel Shell Assembly (4 sheets)

NUH32PTH1-71-1003 Rev 3 NUHOMS 32PTH1 Transportable Canister for PWR Fuel Basket Assembly (8 sheets)

NUH32PTH1-71-1004 Rev 2 NUHOMS 32PTH1 Transportable Canister for PWR Fuel Transition Rails (7 sheets)

NUH32PTH1-71-1010 Rev 2 NUHOMS 32PTH1 Transportable Canister for PWR Fuel Alternate Top Closure (6 sheets)

The staff verified that the drawings include the information described in NUREG-2216 on the (1) materials of construction, (2) dimensions and tolerances, (3) codes, standards, or other specifications for materials, fabrication, examination, and testing (4) welding specifications, including location and nondestructive examination (NDE), (5) coating specifications and other special material treatments that perform a safety function and (6) specifications and requirements for alternative materials.

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

7 2.0 STRUCTURAL AND MATERIALS EVALUATION 2.1 STRUCTURAL EVALUATION The objective of the structural evaluation of the TN Eagle transportation package is to establish that the information summarized in the SAR is adequate to demonstrate compliance with the regulatory requirements established in 10 CFR Part 71 and the supporting analysis is aligned with the procedures and standards for NRC approval of packaging and shipping procedures for fissile material, as applicable.

The TN Eagle transportation packages are designed for the offsite transfer of spent fuel in dry storage containers stored in horizontal storages modules (HSM). The dry storage containers contain BWR and PWR spent fuel. The TN Eagle package is designed to transport DSCs that were designed, fabricated, loaded, and maintained under storage licenses. The packages consist of either a TN Eagle LC (Large Canister) or TN Eagle SC (Standard Canister) cask with the corresponding dry DSC as shown below.

DSC Type TN Eagle cask Type EOS37PTH LC EOS89BTH LC 32PT SC 32PTH1 Type 1 SC 32PTH1 Type 2 SC 24PT4 SC 24PT1 SC FO/FC/FF SC Package

Description:

10 CFR 71.33 requires that the applicant in their SAR include a description of the package in sufficient detail to identify it accurately and provide a basis for its evaluation. This part of the staff review documents the information provided in the SAR to comply with the requirements of 10 CFR 71.33 that are relevant to the structural description of the package.

Package Classification: In complying with 10 CFR 71.33(a)(1), in SAR subsection 1.1.3 the applicant has classified the package as a Type B, Fissile Material Package. The staff finds that this information is adequate to meet the requirements of 10 CFR 71.33(a)(1).

Gross Weight: In complying with the requirements of 10 CFR 71.33(a)(2), the applicant, in SAR table 21 and table 22, presented the computed weights of the different combinations of the package that are addressed in this SAR for TN Eagle LC and TN Eagle SC, respectively. In SAR drawing TNEAGLE011100, the applicant has identified two models of the package as TN Eagle SC and TN Eagle LC. The two configurations are similar, except for the shell thickness of the forged cask body, the design of the shielding rings, and the overall length of the cavity. The same drawings establish the gross weight for TN Eagle SC as 163,000 pounds and that of TN Eagle LC as 164,000 pounds. The staff finds that this information is adequate to meet the requirements of 10 CFR 71.33(a)(2).

Model Number: In complying with 10 CFR 71.33(a)(3), the applicant states that the package has two models which are identified in SAR subsection 1.1.3 as TN Eagle LC and TN Eagle SC. The staff finds that this information is adequate to meet the requirements of 10 CFR 71.33(a)(3).

8 Identification of Containment System: The applicant, in SAR section 1.2.1, describes the containment boundary and the components that makeup the containment system in the subsection titled Closure System. The staff finds that this information is adequate to meet the requirements of 10 CFR 71.33(a)(4).

Materials of Construction, Weights, Dimensions and Fabrication Methods: The receptacles for the radioactive material are the DSCs. SAR section 1.2.3 describes the different DSC types that will be used with the TN Eagle packaging. SAR subsection 1.2.3.1 through 1.2.3.6 describe the various DSCs and their fabrication methods and materials of construction. The staff finds that this information is adequate to meet the requirements of 10 CFR 71.33(a)(5)(i). The internal structure for the radioactive material is the spent fuel basket. SAR subsection 1.2.3.1 through 1.2.3.6 describe the various baskets and their fabrication methods and materials of construction, for the different DSC types that will be used with the TN Eagle packaging. The staff finds that this information is adequate to meet the requirements of 10 CFR 71.33(a)(5)(iii). In SAR section 1.6.3 through 1.6.7 the applicant describes the different DSC and the fuel baskets associated with each. The sections also identify the venting and sampling ports that are on the package and their final configuration post fuel load. The description includes the material of fabrication and the associated design code. The lifting and tiedown devices for the TN Eagle are described in SAR section 2.3.1 and 2.3.2 respectively. The cask will only be lifted in the horizontal position and transported in the same manner. The staff finds that this information is adequate to meet the requirements of 10 CFR 71.33(a)(5)(iv). The maximum normal operating pressure is described in SAR section 3.3.3 in consideration of the different configurations that are possible for the DSC. The staff finds that this information is adequate to meet the requirements of 10 CFR 71.33(b)(5).

Based on these findings, the staff concludes that the applicant has provided adequate description of the proposed package to identify the package accurately and provide a sufficient basis for the structural evaluation of the package in compliance with 10 CFR 71.33.

2.1.1 Description of Structural Components of the Package The cask consists of a forged steel hollow cylinder, which forms the cavity for the DCS, open at the top and with a circular opening at the bottom. The cylinder is forged from a single casting.

The open top is sealed with the primary lid after placement of the DCS. The lid has an orifice access for venting. The access to the venting orifice is sealed with the lid orifice cover plate after completion of venting. The bottom opening provides access to a ram used for loading the DCS into and out of the TN Eagle. On completion of the loading operations the access is sealed by a ram access cover plate.

The containment boundary is defined by the forged cask body, the primary lid and its lid orifice cover plate, the ram access cover plate (RACP), the lid inner seal, orifice cover plate seal, and RACP inner seal for normal conditions of transport (NCT) and hypothetical accident conditions (HAC).

The top edge of the forged cask has a bigger outside diameter, creating a thicker flange that is machined to accommodate the primary lid. The top lid is fixed to the cask body with bolts in holes threaded into the body of the cask.

Shielding rings with resin blocks are shrink-fitted on the steel forged body. The resins provide neutron shielding while the steel of the cask provides gamma shielding. A non-sliding bottom ring confines the rings in place.

9 The cask has two impact limiters (ILs), one at the top and the other at the bottom completing the TN Eagle transportation package. The purpose of the IL is to protect the TN Eagle containment boundary leak tightness from the impacts due to free drops in NCT and HAC. The top and the bottom ILs are of similar design, a stainless-steel adapter serves as the interface between the cask and the energy absorbing material, which are blocks of aluminum honeycomb. A stainless-steel casing with gussets attached to the adapter and encompassed by the aluminum honeycomb make up the IL. A layer of neutron shield resin is placed in the adapter of the IL for additional shielding.

SAR section1.2.1 provides a more detailed description of the general arrangement of the package.

Functions of Structural Components The TN Eagle consists of three major structural components: the cask body, one of several associated dry shielded canisters (DSCs), and the ILs (top and bottom). The cask body is designed to provide protection against environmental loads and loads from normal and hypothetical conditions of transport. Each DSC consists of a metallic shell with a basket assembly to hold the Spent Nuclear Fuel (SNF) and serves as the confinement boundary for the radioactive material. The impact limiters serve as mechanisms to absorb the drop energy that can occur from a package drop under different scenarios. These components are described in chapter 1 of the SAR. SAR section 2.1.1 further describes the structural design for the TN Eagle package. SAR section 1.5 provides drawings and material description for all the relevant components of the TN Eagle package.

Identification of Design Codes and Standards Section 2.1.2 of the application discusses the structural design criteria for the TN Eagle package. The applicant described the cask containment vessel in SAR chapter 4 and states that it is designed to the maximum practical extent as an ASME Class I component in accordance with the rules of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section III, Subsection NB. Section 2.1.2. Additional information is provided in SAR tables 23 and 2-4.

The applicant states that certain components such as the shielding rings, bottom closure plate and its bolts, top handling ring, bottom ring, and closing plate are not part of the containment vessel but do have structural functions. The applicant states that these non-containment structures are designed, fabricated, and inspected in accordance with the ASME Code,Section III, Subsection NF.

Regarding the DSC Shell Assembly, the applicant states that the canisters for the TN Eagle are designed in accordance with the following previously licensed DSCs and are summarized in SAR table 2-3.

EOS37PTH and EOS89BTH licensed in Orano TN, NUHOMS EOS System Updated Final Safety Analysis Report, Revision 3. Docket No. 07201042.

32PTH1, 32 PT, and 24PT4, Orano TN, NUHOMS-MP197 Transportation Package Safety Analysis Report, Revision 20. Docket No. 07109302.

FO/FC/FF and 24PT1 licensed in Orano TN, Safety Analysis Report for the NUHOMS-MP187 Multi-Purpose Cask, Revision 17. Docket No. 07109255.

10 The baskets for all the DSCs are designed, fabricated, and inspected in accordance with the ASME Code,Section III, Subsection NG to the maximum practical extent with the applicable code year and code alternatives specified in section 1.6 of SAR chapter 1 and as further summarized in SAR tables 2-5 and 2-6.

The applicant states that the ILs and its bolts are designed to resist, the applied loads and to prevent separation of the limiters from the cask during an impact. The design criteria of the IL and its bolts are specified in SAR appendix 2.11.3.

The top handling ring and bottom shielding rings are classified as tiedown devices as per 10 CFR 71.45(b) and are designed as such.

General Standards for All Packaging: Section 2.2 of the application discusses the general requirements for all packages. Requirements for minimum package size, tamper-indicating features and positive closure are required to be identified for a package per 10 CFR 71.43.

Minimum Package Size: SAR section 2.2.1 provides the minimum package size. The staff notes that the overall size of the TN Eagle without the ILs to be greater than the 10 cm (4 in.) required by 10 CFR 71.43 (a). The staff finds the package size to comply with the minimum size requirements of 10 CFR 71.43(a).

Tamper-indicating Feature: The access to the TN Eagle package is via the primary lid as well as access through the different orifices, ports, and ram access. During transportation, the IL covers all these entry points. A security wire seal is installed through the guide tubes on both the top and bottom ILs preventing access to the IL bolts. An intact seal demonstrates that there has been no unauthorized entry or purposeful tampering with the package. The staff finds that this tamper-indication feature satisfies the requirements of 10 CFR 71.43 (b).

Positive Closure: All bolts leading to access into the containment vessel have a large preload applied that will protect against any inadvertent opening including loads that can occur during transportation. The torque on these bolts will prevent loosening except deliberate loosening with a wrench. However, access to the bolts is prevented by the tamper-indicating feature of the package. Therefore, the TN Eagle cannot be inadvertently accessed, and any evidence of unauthorized access will be detected. The staff finds that the requirements included in 10 CFR 71.43 (c) for positive closure are met by the large preload applied to the bolts.

Package Valve: The TNEagle does not have any valves or other devices whose failure would allow radioactive material to escape. The staff finds the TN Eagle cask is in compliance with the requirements of 10 CFR 71.43 (e) due to the absence of any other device on the package whose failure has the potential to release radioactive material.

Lifting and Tiedown Standards for All Packages SAR section 2.5 discusses criteria, including design loads, used for lifting and tiedown devices.

The following subsection provides further information and the evaluation performed by the staff.

Lifting Devices: In SAR section 2.3.1 the applicant stated that, for onsite lifting and transfer operations, the TN Eagle will only be lifted in the horizontal orientation. Due to this, the TN Eagle will be lifted using two slings, one around the top flange of the forged cask body and the other around the bottom ring. The applicant discusses this analysis in SAR appendices 2.11.1 and 2.11.9. Appendix 2.11.1 discusses the TN Eagle NCT evaluation. In this appendix, the

11 applicant states that the weight of the ILs is not included for the lifting operation for which slings are used to lift the cask onto the transport frame. The applicant has justified not including the weight of the ILs in this operation as the lift is conducted without the ILs under the heavy lift program of an NRC regulated facility, with the lift height being controlled by the acceptable risk for the facility. The staff determined that it is acceptable to not include the IL weights in the lift analysis along the top flange of the forged cask and around the bottom ring, which are structural parts of the package. SAR subsection 2.11.9.7.3 identifies the lifting stress in the bottom ring and SAR table 2.22.9-5 provides a summary of stresses. The staffs review of this information finds that the factor of safety is greater than that required per 10 CFR 71.45 (a), and concludes that the requirements of 10 CFR 71.45 (a) for lifting devices are met.

Tie-Down Devices: During transportation, the TN Eagle package will be confined in a transport frame with a shear key to resist movement. In SAR section 2.3.2, the applicant states how the different features of the transport frame accommodate the transportation loads. To resist movement due to longitudinal acceleration, the frame includes a shear key that sets into the top handling ring of the TN Eagle package. Saddles at each end of the frame restrain the TN Eagle package against vertical and lateral accelerations. The top saddle latches with the top handling ring and the rear saddle latches with the shielding ring at the bottom end of the cask. The applicant analyzed both the handling ring and shielding ring for the loads defined by 10 CFR 71.45 (b). In appendix 2.11.9 subsection 2.11.9.6.3 the applicant presents the radial stress and the longitudinal stress as a result of the tie-down loads. From a review of these stresses and their comparison to the allowable stresses, the staff finds that the stresses experienced by these components are less than the yield stress. Based on this finding, the staff concludes that the components meet the requirements set forth by 10 CFR 71.45 (b).

2.1.2 General Considerations for Structural Evaluation The structural evaluation of the TN Eagle package and its contents under NCT and HAC uses the guidance in NRC Regulatory Guide 7.6, Revision 1, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels, 1987 and the NRC Regulatory Guide 7.8, Revision 1, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material, 1989. SAR table 2.11.1-3 lists the load combinations used in the analysis. The design acceptance is based on ASME codes as applicable. In addition, the applicant has relied on the approved licenses for: Orano TN, NUHOMS EOS System Updated Final Safety Analysis Report, Revision 3. Docket No. 07201042; Orano TN, NUHOMS-MP197 Transportation Package Safety Analysis Report, Revision 20. Docket No.07109302 and Orano TN, Safety Analysis Report for the NUHOMS-MP187 Multipurpose Cask, Revision 17. Docket No. 07109255.

Evaluation by Analysis: The evaluation of the structural performance of the TN Eagle package is solely based on finite element modeling and analytical solutions using this model. The structural evaluation of the TN Eagle package is performed using finite element codes ANSYS and LSDYNA. Stress analysis with allowable stress criteria is used with ANSYS finite element models while the LSDYNA finite element code is utilized for non-linear analysis considering deformation beyond elastic limits.

Description of Finite Element Models for Analysis TN Eagle Model for NCT Model

Description:

For the TN Eagle NCT analysis, a 3D finite element model (FEM) was developed for the analysis performed using the ANSYS code. The model represents a 180-

12 degree symmetric section of the TN Eagle structural components, which includes the forged cask body, primary lid and primary lid bolts, as shown in SAR figure 2.11.1-1. The components are modeled using solid elements with linear elastic properties. SAR chapter 7 identifies the material properties used in the model.

Acceptance Criteria: The acceptance criteria is based on the design stress determination process defined in ASME,Section III, Subsection NB design rules. SAR table 2.11.1-1 identifies the criteria applicable to each component, and SAR table 2.11.1-2 establishes the stress allowable applicable to each component. The load combinations used in the NCT design are shown in SAR table 2.11.1-3. A screening criterion is used to determine the acceptability of a load combination meeting the acceptable stress limits when considering a combination of maximum primary membrane stress, primary membrane stress and bending stress, and secondary membrane stress and bending stress. When the stress results are acceptable against the applicable allowable stresses, the stresses are deemed acceptable, and the screening process is ended.

Loading: The weight of the contents of the TN Eagle is applied on the inner surface of the forged cask body. The weight used in the analysis bounds the maximum weight of the DSC that can fit into the TN Eagle. The IL weights, as calculated in SAR section 2.1.1.4, are included in the model. The accelerations in the axial direction to the forged body and the weight of the IL are applied along with the reactions from the 3 g lifting. The forged body and primary lid are evaluated at two thermal loading conditions, hot and cold. Bolt preload of 500 kN is applied to the lid bolts. Fabrication stresses from shrink fitting of the rings onto the forged cask are simulated using pressure around the entire circumference of the forged body, from the bottom edge to the flange as computed in SAR appendix 2.11.9. The differential thermal expansion is simulated by applying an external inward pressure, as computed in SAR appendix 2.11.9, and is considered as a secondary load. External pressure load is applied at all external nodes of the forged body, while the internal pressure is applied at the internal surface nodes of the forged body. The loads from acceleration during transportation are added directly to the body of the forged cask as reactions at the points of attachment. Stresses resulting from the free fall are combined with other loads cases and assessed against the NCT stress criteria. The computation of the loads from the 1 ft free drop are addressed in SAR appendix 2.11.3.

TN Eagle Model-Free Drop Analysis for HAC Model

Description:

For the TN Eagle cask HAC drop analysis, a 3D finite element model was developed for the analysis performed using the LSDYNA code. The model represents a 180-degree symmetric sector of the TN Eagle cask, including the structural components identified in SAR table 2.11.7-1. The FEM was developed with and without the IL as shown in SAR figures 2.11.7-1 and 2.11.7-2, respectively. Parts of the cask are modeled using solid fully integrated elements with linear elastic material models. The material properties are shown in SAR table 2.11.7-1. Weights of components not explicitly modeled are included in the model by adjusting the mass of the modeled parts.

The ILs are modeled in detail to include the different components of the IL. These components consist of the adapter, aluminum honeycomb blocks (AH) as shown in SAR figure 2.11.7-4, the metallic shell, IL bolts and bolt guides. The adapter, shell, bolt guides and bolts are modeled with solid fully integrated elements and using the linear plasticity model properties from SAR table 2.11.7-5. The AH blocks in the IL are modeled using fully integrated solid elements and using the material properties that reflect the oriented strength properties of the stacked corrugated aluminum sheets used in the manufacture of the AH blocks. The properties of the AH block are identified using a series of curves as shown in SAR table 2.11.7-6. The LCA curve

13 defines the dependency of the crush stress with respect to the grain angle as shown in SAR table 2.11.7-7. The LCB curve defines the behavior of the AH in the weak axis as shown in SAR table 2.11.7-8. The LCC curve defines the behavior of the AH in the strong axis and is as shown in SAR table 2.11.7-9. The LCAB, LCBC and LCCA curves defines the shear damage of the AH material as described in SAR tables 2.11.7-10 and 2.11.7-11. The AH material model has been benchmarked using test data in Orano TN Document TR-1004453, Rev. 1, Benchmarking of the Constitutive Model of Biaxial Aluminum Honeycomb.

Contacts: The area of contact between the IL and the cask is modeled using the capabilities in LS-DYNA to model contact surfaces. Hence the dynamic friction between the two components is captured using these features of LS-DYNA. To simulate the connection of the ILs with the cask body, the nodes of the shanks of the bolts are tied to the cask at the appropriate locations.

Loading: The following drop events are analyzed using this model:

1.

1 ft Side Drop 2.

30 ft End Drop on primary lid end 3.

30 ft End Drop on bottom end 4.

30 ft Side Drop 5.

30 ft Center-of-Gravity (CG)-over-Corner Drop on primary lid end 6.

30 ft Center-of-Gravity (CG)-over-Corner Drop on bottom end 7.

30 ft 100 Slap Down on primary lid end 8.

30 ft 100 Slap-down on bottom end 9.

30 ft 200 Slap down on primary lid end

10. 30 ft 200 Slap down on bottom end The events are considered for both the hot and cold conditions.

Model Setup: Symmetric boundary conditions are applied with degrees of freedom in X-translation, Y and Z in rotation. SAR figure 2.11.3-2 provided the initial conditions. The impact surface is modeled using the LS-DYNA rigid wall element. Bolt pre-loads are applied to primary lid bolts, bottom closure plate (BCP) bolts, RACP bolts and IL bolts. The pre-loads are applied incrementally at the interface of the bolt heads and washers. Shrink fit loads are applied to the forged cask body and the BCP using the shielding rings, closing plate, bottom ring and the top handling ring. The load is applied to the inner surface of the above part incrementally.

Data Processing: The rigid body accelerations are derived from the maximum impact force for the end, side and CG-over-corner drops following the guidance in NUREG/CR-3966, Methods for Impact Analysis of Shipping Containers, 1987. Accelerations for slap down drops are extracted from the LS-DYNA accelerometers located at the top and bottom. Accelerations for other than slap down drops are derived from the LS-DYNA generated Administrative Services Center II files. Filtering of the resultant accelerations is used to separate the high frequency natural vibration accelerations from the low frequency rigid body accelerations. A low order Butterworth filter is used for this purpose.

Acceptance Criteria: Is based on NRC Regulatory Guide 7.6, Rev. 1, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels, 1978, which indicates that the performance of the containment boundary and other principal shells of the cask remain linear elastic. The applicant has adopted the linear elastic properties of the cask material as the bases of design limits for the cask design.

Condition in CoC: Staff requested the applicant to include the transportation frame in the

14 numerical model of the TN Eagle package for drop analyses conducted under HAC to obtain assurance that the response of the impact limiters remains uninfluenced by the presence of the transportation frame. The current analysis did not include the potential influence of the frame on the response of the package: as such, the frame may impact the rigid surface prior to the engagement of the impact limiters and if the frame is considered, the deformation of the impact limiters may be constrained. The applicant did not respond at this time and the CoC includes the condition that lifting the package, while attached to the transport frame, is not authorized. The TN Eagle package is handled solely in a horizontal position.

Evaluation by Test: This evaluation of the TN Eagle does not use testing as a basis for demonstrating the structural integrity of the package.

2.1.3 Normal Conditions of Transport In SAR section 2.5, the applicant addresses the regulatory requirements for NCT and how these requirements are met by the designed features of the package. The following safety evaluation report (SER) subsections present the staffs evaluation of the NCT requirements in 10 CFR 71.71.

Heat: The applicant refers to chapter 3 of the SAR for a detailed description of the thermal analyses. SAR table 3-8 summarizes maximum temperatures calculated for the TN Eagle subjected to hot environment conditions, which were found to be consistent with the ambient temperature specified in 10 CFR 71.71(c)(1). Temperatures from the analyses in chapter 3 are applied to finite element (FE) models in appendix 2.11.1 for the estimation of thermal stresses.

The thermal expansion evaluations of the TN Eagle cask cavity and DSCs in both the radial and axial directions are described in appendix 2.11.14. The differential thermal expansion of the cask and the outer rings is addressed in appendix 2.11.9. The applicant determined, based on their analysis, that there is adequate clearance between the various components of the DSC and cask to allow free thermal expansion. Consequently, the staff finds that no significant stress will develop between the cask and DSC from thermal expansion. Based on this finding, the staff concludes that the requirements of 10 CFR 71.71(c)(1) are satisfied.

Cold: The applicant refers to SAR chapter 3 and appendix 2.11.1 for a detailed description of thermal analyses performed for the TN Eagle subjected to cold environment conditions. The analyzed temperatures were found to be consistent with the ambient temperature specified in 10 CFR 71.71(c)(2). The applicant determined that there is adequate clearance between the various components of the DSC and cask to allow free thermal contraction. Consequently, the staff finds that no significant stress will develop in the TN Eagle due to thermal expansion of the DSCs. Based on this finding, the staff concludes that the requirements of 10 CFR 71.71(c)(2) are satisfied.

Reduced External Pressure: The applicant in SAR table 3-12 establishes that the maximum normal operating pressure (MNOP) as 12.1 psig (= 83.5 kPa). The greatest possible pressure difference is computed as 159.9 kPa (= 83.5 kPa + 101.325 kPa-25 kPa) outwards. The applicant in their analysis summarized in SAR appendix 2.11.1 has used a bounding outward pressure of 210 kPa applied to the inner surfaces of the cask cavity. Staff review of this analysis finds that the pressure used in the analysis is greater than the possible pressure differential.

Thus, the staff finds that the results of the analysis bound the regulatory requirements for reduced external pressure. Based on this finding, the staff concludes that the requirements of 10 CFR 71.71(c)(3) are satisfied.

Increased External Pressure: 10 CFR 71.71(c)(4) requires the package to be subject to an

15 externally applied pressure of 140 kPa. The applicant in appendix 2.11.1 analyzed the cask for a bounding inwards pressure of 175 kPa applied at the outer surfaces of the cask keeping the internal pressure at 0 kPa. The staff finds that this analysis condition bounds the regulatory requirements for increased external pressure. Based on this finding, the staff concludes that the requirements of 10 CFR 71.71(c)(4) are satisfied.

Vibration and Fatigue: The applicant applied the guidance in NRC, NUREG 766510, Shock and Vibration Environments for Large Shipping Containers on Rail Cars and Trucks,1977, and applied the recommended accelerations in three orthogonal directions at the center-of-gravity of the package. The resulting reactions from these accelerations as computed in Design Calculation TNE01-0209, Revision 0, are applied to the cask model at the interfaces between the saddles, latches, and the shear key. The applicant used these reactions in the structural analyses for the containment boundary in SAR appendix 2.11.1, as well as when combining with other load cases as shown in SAR table 2.11.1-3. They were also used in the analyses of the primary lid and ram access cover plate bolts in appendix 2.11.4.9, in considering fatigue in these bolts. The staff reviewed the excerpts provided from the calculation to demonstrate the manual calculation process that was used to derive the reactive forces resulting for the acceleration due to transportation vibration and shock. The staff finds the approach to determine the reactive forces reasonable for quantifying the reactions at the points of attachment between the package and the transportation frame and the inclusion of the reaction in the analysis of the TN Eagle package. Based on this finding, the staff concludes that the requirements of 10 CFR 71.71(c)(5) are satisfied.

The applicant performed a fatigue analysis considering the combined stress from vibration, pressure, and temperature loads in SAR appendix 2.11.8 for the containment boundary and in appendix 2.11.4 for the primary lid and ram access cover plate bolts. The fatigue analysis is based on section XIII-3520 and Mandatory Appendix I of ASME B&PV Code,Section III, Appendices, 2017, with the objective to demonstrate that the containment boundary stresses are within the acceptable NCT fatigue limits. The maximum stress intensities for all individual loads are combined simultaneously for the following sequence of events: Bolt preload, deadweight, pressure fluctuations, temperature fluctuations, vibration, shock and the 1-foot drop. The staff reviewed the information presented in SAR subsection 211.8 and finds that the number of round trips for which the TN Eagle package meets the fatigue stress limits under NCT conditions is 400. Based on this finding, the staff concludes that the TN Eagle meets the fatigue stress limits set by the design acceptance criteria.

Water Spray: SAR section 2.5.6 identifies that various impermeable fabrication materials of the TN Eagle packaging make the exterior of the package spray resistant. In reviewing this information, the staff finds that water spray is likely to have no points of entry to cause any damage to the packaging material. Based on its findings, the staff concludes that water spray will have minimal effect on the TN Eagle package. Based on this finding, the staff concludes that the requirements of 10 CFR 71.71(c)(6) are satisfied.

Free Drop Side Drop Analysis Results: The TN Eagle is always at a horizontal position during handling.

Therefore, only the 1 ft side drop is considered as a credible event. For this event, the ILs are attached. The 1ft side drop is not a credible event for the primary lid and the RACP bolts. The package is analyzed in LS-DYNA described in detail in SAR appendix 2.11.3. The results of the drop analysis are combined with other NCT loads cases as shown in SAR table 2.11.1-3. The stress assessment for the NCT conditions is presented in SAR table 2.11.1-4 and table 2.11.1-

5. The staff reviewed the information in the identified sections and finds that the demand stress

16 in the TN Eagle cask subject to the NCT loadings meets the load combinations of NRC, Regulatory Guide 7.8, Revision 1, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material, 1989 and the stress criteria of ASME B&PV Code,Section III, Division 1, Subsection NB, 2017.

Corner Drop: The applicant in SAR section 2.5.8 states that the corner drop test in accordance with 10 CFR 71.71(c)(8) does not apply to the TN Eagle package as the gross weight of the package exceeds 50 kg (110 lbs.). The staff review finds that this assessment is appropriate and in compliance with the requirements of 10 CFR 71.71(c)(8).

Compression: The applicant in SAR section 2.5.9 states that the compression test does not apply to the TN Eagle package as the mass of the package is in excess of 5,000 kg. The staff review finds that this assessment is appropriate and in compliance with the requirements of 10 CFR 71.71(c)(9).

Penetration: The applicant in SAR section 2.5.10 states that due to the lack of sensitive external protuberances, the 1 meter drop of a 6 kg steel cylinder of 32 mm diameter with a hemispherical head, in accordance with 10 CFR 71.71(c)(10), is of negligible consequence to the TN Eagle package. The staff review finds that this assessment is appropriate and complies with the requirements of 10 CFR 71.71(c)(10).

Primary Lid and Ram Access Cover Plate Bolts: In SAR appendix 2.11.4 the applicant presents the analysis of lid bolts. The bolts are analyzed following the guidance in NRC NUREG/CR-6007, Stress Analysis of Closure Bolts for Shipping Casks, 1992. The staff reviewed the information presented in SAR appendix 2.11.4 for the methodology and the loads considered in the analysis, the materials used, the load combinations along with the gasket seating and prying effects. The staff finds that the individual loads acting on the closure bolts of the primary lid are summarized in SAR table 2.11.4-5; the load combinations are presented in SAR table 2.11.4-6 along with the allowable stress in SAR table 2.11.4-2. The various calculated stress is presented in SAR table 2.11.4-10. A fatigue analysis was considered for the 400 round trips after which the bolts will be replaced. SAR subsection 2.11.4.4 presents the analysis of the ram access cover plate bolts. These bolts are analyzed along the same lines as the primary bolts with bolt allowable stresses shown in SAR table 2.11.4-2 and the computed stresses in SAR table 2.11.4-10. Based on a review of the provided information the staff finds that under NCT conditions the bolts meet the stress acceptance criteria of NUREG /CR-6007, there is a net compressive force in the bolts under all conditions, the bolts have adequate engagement length and demonstrate no-failure in the 400 round trip fatigue cycle.

DSC Shell Assembly: The EOS DSC shell assemblies are qualified previously. In SAR section 2.11.11 the applicant reconciles their use with TN Eagle packages. The shells are analyzed for NCT 1 ft side drop and two thermal loading conditions. The analysis is performed in ANSYS.

The stress acceptance criteria are based on ASME,Section III, Subsection NB requirements, and are presented in SAR tables 2.11.11-1 to table 2.11.11-3. The load combinations are shown in SAR table 2.11.11-6. A triaxial stress check was performed to ensure that no triaxial tensile stress failure occurs for the considered load combinations. The results of the side drop with internal pressure or external pressure are presented in SAR tables 2.11.11-9 and 2.11.11-10 respectively. The thermal stresses for hot and cold conditions are presented in SAR table 2.11.11-7 and table 2.11.11-8, respectively. A comparison between the stress demand and their allowable stress for the weld between the lifting lug and the DSC shell is shown in SAR table 2.11.11-3. A summary of the stress results for the NCT load combinations involving side drop, pressure and thermal are shown in SAR table 2.11.11-11 and table 2.11.11-12. Based on a review of the information in the SAR the staff finds that the EOS shell assemblies meet the

17 acceptance criteria of ASME,Section III.

EOS Baskets: In SAR appendix 2.11.12 the applicant provides information on the reconciliation of the EOS baskets under the NCT conditions for the TN Eagle package. The FE model for EOS-37PTH and EOS-89BTH replicate the information in SAR chapter 1 drawings for the LC configuration. ANSYS is the analytical code used for the analysis along with dynamic load factors (DLF) developed in SAR appendix 2.11.6. Model sections used in the analysis are shown in SAR figures 2.11.12.-1 through figure 2.11.12-4. The mass of the poison plates and aluminum plates are included by increasing the density of the steel grid plates. The continuous support of the basket by the transition rails over the entire length are modeled. Gap elements are used to capture the basket response between the DC and the transportation cask. SAR chapter 1 fuel data provides the fuel mass for the analysis model. For the side loads from a NCT 1 ft side drop the results from SAR appendix 2.11.3 are used with the DLF from SAR appendix 2.11.6 to compute the acceleration imposed on the basket. A similar analysis is performed for the thermal conditions. A bounding value is used in the basket analysis. The allowable stresses are based on ASME Section III, Division 1, Subsection NG. and are summarized in SAR table 2.11.12-1. Allowable stresses for the treaded fasteners connecting the transition rails to the basket grid are shown in SAR table 2.11.12-2 based on NG-3230. Due to the high bounding value used to qualify the basket, a limit state analysis was utilized to justify the high stresses at localized spots for the side drop analysis, which used the provisions of NG-3228 and note nine of Figure NG-3221-1. To further demonstrate that the results do not exceed two-thirds of the yield, as required for such analysis, a 1.5 times higher value was used in a subsequent analysis.

The results of the analysis in the form of maximum stress intensities are presented in SAR table 2.11.12-7. Based on a review of the above information the staff finds that the EOS baskets comply with the ASME acceptance stress levels against all loading conditions. Where the basic primary stresses are exceeded at local hots spots in the basket a limit state analysis shows that the basket has a safety factor of 1.5 against collapse.

Non-EOS Content Under NCT: The Non-EOS Content consists of the 32PT, 32PHT1, and 24PT4 canisters and baskets licensed in Docket No. 07109302 and FO, FC, FF and 24PT1 canisters and baskets licensed in Docket No. 07109255.

32PTH1 DSC Side Drop: In SAR appendix subsection 2.11.10.1.1 the applicant addresses the response of the 32PTH1 canister when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 in conjunction with the dynamic properties of the canister from the original license and the drop acceleration from SAR table 2.11.3-1 the maximum acceleration of the canister in the TN Eagle package was computed as 17.5 g. In the original license, the canister was evaluated for an acceleration of 30 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the canister when used in the TN Eagle package is bounded by the analysis of the original license.

32PT DSC Side Drop: In SAR appendix subsection 2.11.10.1.3 the applicant addresses the response of the 32PT canister when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 in conjunction with the dynamic properties of the canister from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the canister in the TN Eagle package was computed as 15.9 g. In the original license, the canister was evaluated for an acceleration of 25 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the canister when used in the TN Eagle package is bounded by the analysis of the original license.

18 24PT4 DSC Side Drop: In SAR appendix subsection 2.11.10.4 the applicant addresses the response of the 24PT4 canister when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 in conjunction with the dynamic properties of the canister from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the canister in the TN Eagle package was computed as 17.49 g. In the original license, the 24PT4 canister was evaluated for an acceleration of 25 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the canister when used in the TN Eagle package is bounded by the analysis of the original license.

FO, FC, FF and 24PT1 Side Drop: These canisters are not credited in the safety evaluation of the package, hence are not included in the analysis. The staffs review confirms this and finds this acceptable.

32PTH1 Basket Side Drop: In SAR appendix subsection 2.11.10.2.1 the applicant addresses the response of the 32PTH1 basket when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the basket in the TN Eagle package was computed as 17.0 g. In the original license, the basket was evaluated for an acceleration of 23.3 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the basket when used in the TN Eagle package is bounded by the analysis of the original license.

32PT Basket Side Drop: In SAR appendix subsection 2.11.10.2.2 the applicant addresses the response of the 32PT basket when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the basket in the TN Eagle package was computed as 16.8 g. In the original license, the basket was evaluated for an acceleration of 25.0 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the basket when used in the TN Eagle package is bounded by the analysis of the original license.

24PT4 Basket Side Drop: Due to the configuration of this basket, the individual components of the basket needed to be evaluated. Assessments were performed for the spacer disk, guide sleeve and spacer rod.

Spacer Disk: In SAR appendix subsection 2.11.10.2.3 the applicant addresses the response of the 24PT4 basket spacer disk when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket spacer disk from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the basket spacer disk in the TN Eagle package was computed as 16.85 g. In the original license, the basket was evaluated for an acceleration of 21.0 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the basket spacer disk when used in the TN Eagle package is bounded by the analysis of the original license.

Guide Sleeve: In SAR appendix subsection 2.11.10.2.3 the applicant addresses the response of the 24PT4 basket guide sleeve when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket guide disk from the original license and the drop acceleration

19 from SAR table 2.11.3-1, the maximum acceleration of the basket guide disk in the TN Eagle package was computed as 16.4 g. In the original license, the basket was evaluated for an acceleration of 27.0 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the basket guide sleeve when used in the TN Eagle package is bounded by the analysis of the original license.

Spacer Rod: In SAR appendix subsection 2.11.10.2.3 the applicant addresses the response of the 24PT4 basket spacer rod when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket spacer rod from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the basket spacer rod in the TN Eagle package was computed as 16.4 g. In the original license, the basket was evaluated for an acceleration of 30.0 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the basket spacer rod when used in the TN Eagle package is bounded by the analysis of the original license.

FO/FC Side Drop: Due to the configuration of this basket, the individual components of the basket needed to be evaluated. Assessments were performed for the spacer disk, guide sleeve and spacer rod.

Spacer Disk: In SAR appendix subsection 2.11.10.2.4 the applicant addresses the response of the FO/FC basket spacer disk when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket spacer disk from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the basket spacer disk in the TN Eagle package was computed as 16.85 g. In the original license, the basket was evaluated for an acceleration of 25.1 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the basket spacer disk when used in the TN Eagle package is bounded by the analysis of the original license.

Guide Sleeve: In SAR Appendix subsection 2.11.10.2.4 the applicant addresses the response of the FO/FC basket guide sleeve when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket guide sleeve from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the basket guide sleeve in the TN Eagle package was computed as 16.4 g. In the original license, the basket was evaluated for an acceleration of 25.1 g which bounds the acceleration in the TN Eagle package.

Based on a review of this analysis, the staff finds that the basket guide sleeve when used in the TN Eagle package is bounded by the analysis of the original license.

Spacer Rod: In SAR appendix subsection 2.11.10.2.4 the applicant addresses the response of the FO/FC basket spacer rod when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket spacer rod from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the basket spacer rod in the TN Eagle package was computed as 16.4 g. In the original license, the basket was evaluated for an acceleration of 30.0 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the basket spacer rod when used in the TN Eagle package is bounded by the analysis of the original license.

FF Side Drop: Due to the configuration of this basket, the individual components of the basket needed to be evaluated. Assessments were performed for the spacer disk, fuel canister and

20 support plate.

Spacer Disk: In SAR appendix subsection 2.11.10.2.5 the applicant addresses the response of the FF basket spacer disk when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket spacer rod from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the basket spacer disk in the TN Eagle package was computed as 17.2 g. In the original license, the basket spacer disk was evaluated for an acceleration of 25.1 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the basket spacer disk when used in the TN Eagle package is bounded by the analysis of the original license.

Fuel Canisters: In SAR appendix subsection 2.11.10.2.5 the applicant addresses the response of the FF basket fuel canisters when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket fuel canisters from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the basket fuel canisters in the TN Eagle package was computed as 16.4 g. In the original license, the basket fuel canister was evaluated for an acceleration of 25.1 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the basket fuel canisters when used in the TN Eagle package is bounded by the analysis of the original license.

Support Plate: In SAR appendix subsection 2.11.10.2.5 the applicant addresses the response of the FF basket support plates when subject to the side drop accelerations resulting from its use in the TN Eagle package. Using the DLF computed in SAR appendix 2.11.6 with the dynamic properties of the basket support plates from the original license and the drop acceleration from SAR table 2.11.3-1, the maximum acceleration of the basket support plates in the TN Eagle package was computed as 16.4 g. In the original license, the basket fuel canister was evaluated for an acceleration of 24.82 g which bounds the acceleration in the TN Eagle package. Based on a review of this analysis, the staff finds that the basket support plates when used in the TN Eagle package is bounded by the analysis of the original license.

24PT1 Side Drop: The analysis of the components of the 24PT1 basket is bounded by the original analysis for the FO/FC configuration. Hence, no further evaluation of its structural adequacy is needed for its use in the TN Eagle transportation package configuration.

Based on their review findings the staff concludes that the TN Eagle transportation package complies with the requirements of 10 CFR 71.71.

2.1.4 Hypothetical Accident Conditions SAR section 2.6 discusses the response of the TN Eagle package to HAC as prescribed in 10 CFR 71.73. The design criteria used in the analysis are described in SAR section 2.1.2.1 to ensure that the requirements of 10 CFR 71.51 are satisfied.

9 m (30 ft) Free Drop-Containment Boundary: SAR appendix 2.11.3 presents the details of the drop analysis. The drop simulations are conducted using the FEM of the cask body and the IL for the different orientations evaluated in this SER. The applicant evaluated the response of the TN Eagle package for a free drop from a height of 30 feet onto an unyielding surface. The rigid body accelerations of the TN Eagle are determined in the LS-DYNA analyses presented in appendix 2.11.3. The staff reviewed the additional information provided in response to requests for additional information (RAIs) and finds that the methodology used to benchmark the biaxial

21 aluminum honeycomb as an energy absorbing material for the TN Eagle impact limiters is acceptable. In addition, the staff finds that the modeling of the cask body as an isotropic linear elastic material is reasonable as the deformation in the AH blocks of the IL dissipates a significant part of the drop energy. Energy plots of the drops are shown in SAR appendix figures 2.11.3-41 through 51. SAR appendix 2.11.3 presents the structural evaluation of the TN Eagle containment boundary for the drops required for compliance with 10 CFR 71.73. and the load combinations of NRC, Regulatory Guide 7.8, Rev.1, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive material, 1989. The stresses allowable for the forged cask body, primary lid and the RACP are shown in SAR table 2.11.2-2 and the stress results are shown in SAR table 2.11.2-3. The staffs review of the presented information finds that for the cask forged body, primary lid and the RACP, the design margins under all the load conditions are positive. The staff concludes that this meets the design code requirements for the containment boundary performance.

Primary Lid and Ram Access Cover Plate Bolt Stress: The primary lid and RACP cover bolts maintain a leak tight seal during the HAC events. In SAR appendix 2.11.4 the applicant has analyzed the ability of the bolts to perform this function along with the stresses in the bolt threads and a fatigue assessment under repeated use. The stress analysis is performed in accordance with the guidance in NUREG/CR-6007, Stress Analysis of Closure Bolts for Shipping Casks, 1992. In the stress analysis of the primary lid bolts the applicant has considered bolt preload, gasket seating load, internal pressure load, temperature load, impact, puncture, and external pressure load. The individual loads on the primary lid bolts are listed in SAR appendix table 2.11.4-5 and the RACP bolt loads are listed in table 2.11.4-7. Tables 2.11.4-6 and 2.11.4-8 present the load combinations for the primary lid cover bolts and the RACP bolts, respectively. The resulting bolt stresses along with their allowable stresses are shown in SAR appendix table 2.11.4-10. The fatigue analysis provides a quantitative estimate of the fatigue damage to the bolts from repeated use of the bolts. This is conducted using a damage factor for the NCT loads. The staffs review of the presented information finds that the stress ratio for the primary lid and the RACP bolts is considerably less than one. Based on this review the staff concludes that the bolts of the primary lid and RACP cover bolts maintain a leak tight seal during the HAC events.

DSC Shell and Basket Assemblies Under HAC: The DSC shells are analyzed for inertial loads developed by the free fall and internal and external pressures. A set of DLF are used to amplify the inertial loads selected using natural frequency of the component from the DLF computed in SAR appendix 2.11.6. The results of the stress analysis for the DSC canisters are shown in SAR appendices 2.11.10 and 2.11.12. Based on a review of the analysis presented the staff finds that the canisters and basket assemblies are adequate for the applied HAC loads and meet the design requirements of the regulatory guidance documents and ASME B&PV Code,Section III, Division 1, Subsection NG and Appendices.

Fuel Assemblies: Fuel assemblies are not evaluated for structural integrity under HAC conditions as they are considered reconfigured under HAC and evaluated for this reconfigured condition for thermal, shielding and criticality in the appropriate SER sections. The staff concludes that the free drop tests, in aggregate, satisfy the requirements of 10 CFR 71.73(c)(1).

Crush: The TN Eagle package weighs more than 5000 kg which is more than the weight of packages that require the crush test under 10 CFR 71.71(c)(2). Therefore, the staff finds that the dynamic crush test of 10 CFR 71.73(c)(2) does not apply to the TN Eagle package.

Puncture: SAR section 2.63 discusses the puncture test for the TN Eagle package. The applicant concluded that sufficient margin of safety exists against bending and shear for the

22 forged cask body and the primary lid of the TN Eagle. The thickness of the forged cask body, primary lid and the ram access cover are adequate to resist a puncture on a mild steel bar. SAR appendix 2.11.5 provides a detailed evaluation of the puncture drops. The staff reviewed the analysis in appendix 2.11.5 and finds that there is a positive margin for both bending and shear stress due to puncture for the cask body, primary lid and the RACP. The impact on the shielding ring has a negative margin against shear due to puncture. This means that the shielding ring is punctured but the containment boundary is maintained by the unaffected cask wall. The reduction in the shielding capability as a result of this puncture is included in the shielding analysis. Based on its review findings, the staff concludes that the TN Eagle package meets the requirements of 10 CFR 71.73(c)(3). In appendix 2.11.5.1.5, the applicant provides a qualitative discussion on the effects of puncture on the shielding ring and states that local damage may occur after such an accident condition. This may result in the loss of localized shielding capabilities. The applicant has considered two scenarios as a result of such a damage in the shielding calculations.

Thermal: In SAR appendix subsection 2.11.2.2.2 the applicant presents information about the treatment of the fire accident load on the TN Eagle containment boundary at postfire steady state conditions 30 minutes after the fire. An internal pressure is applied on the inner surface of the containment boundary which bounds the internal pressure from the thermal calculation in SAR section 3.4.4.1. The material properties used are consistent with SAR table 3-16 temperatures. The thermal stresses are not included as they are considered as secondary stresses. The stress results of load combinations including the effects of fire are shown in SAR table 2.11.2-3. Based on a review of this information the staff finds that the applicant has adequately assessed the effects of the fire on the integrity of the containment boundary consistent with the requirements of 10 CFR 71.73(c)(4).

Immersion - Fissile Material: In SAR subsection 2.4.6.5 the applicant presents information about the treatment of the effect of immersion in 3 feet of water for leakage post drop accident on the TN Eagle containment boundary. SAR appendix 2.11.2.2.3 evaluates the TN Eagle containment boundary for immersion under 15 m of water by applying an external pressure equivalent to this head of water, which bounds the pressure from the requirements of 10 CFR 71.73(c)(6). The stress results of load combinations including the effect of immersion is shown in SAR table 2.11.2-3. Based on a review of this information the staff finds that the applicant has adequately assessed the effects of immersion on the integrity of the containment boundary consistent with the requirements of 10 CFR 71.73(c)(5).

Immersion - All Packages: The staff agrees with the applicant that this requirement is met by subjecting the TN Eagle package to the external pressure of 15 m of water to satisfy the requirements of 10 CFR 71.73(c)(5). The staff finds that this bounding evaluation satisfies the requirements of 10 CFR 71.71(c)(6). Based on the findings of the staffs review, the staff concludes that the TN Eagle package design has demonstrated ability to meet the integrated requirements of 10 CFR 71.73.

Based on review of the statements and representations in the application, the staff concludes that the structural design has been adequately described and analyzed. Therefore, the staff concludes that the TN Eagle package has adequate structural integrity to meet the requirements of 10 CFR Part 71.

2.2 MATERIALS EVALUATION

23 The staff evaluated the material characteristics of the TN Eagle design for transportation of DSCs, with a transport index for the package of greater than 10 and shipped by exclusive conveyance.

2.2.1 Drawings The applicant provided drawings for the package including details of the outer package and the inner package for the TN Eagle transportation package. The applicants drawings included component safety classification, a bill of materials with material specifications for each component, and dimensions of the components with tolerances.

The applicant also provided drawings for the DSC designs that will be transported in the TN Eagle transportation package. These include the EOS-37PTH and EOS-89BTH DSCs included in the storage CoC No. 1042, the 24PT1 and the 24PT4 DSCs included in the storage CoC No.

1029, the 32PT and the 32PTH1 DSCs included in the storage CoC No. 1004, and the fuel only (FO), fuel and control components (FC) and failed fuel (FF) DSCs included in special nuclear material license (SNM) -2510. The applicants drawings for each DSC included component safety classification, a bill of materials with material specifications for each component, and dimensions of the components with tolerances.

The staff reviewed the drawings using the guidance in NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material, issued August 2020, NUREG/CR-5502, Engineering Drawings for 10 CFR Part 71 Package Approvals, issued May 1999, and Regulatory Guide 7.9, Standard Format and Content of Part 71 Applications for Approval of Packages for Radioactive Material, for the recommended content of engineering drawings. In addition, the staff used NUREG/CR-6407, Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety, issued February 1996, and NRC Regulatory Guide 7.10, "Establishing Quality Assurance Programs for Packaging Used in the Transport of Radioactive Material," Appendix A, A Graded Approach to Developing Quality Assurance Programs for Packaging Radioactive Material, for guidance on safety classification of transportation packaging components.

The staff verified that the drawings included design features considered in the package evaluation, including:

the containment system closure device internal supporting or positioning structures gamma shielding outer packaging heat-transfer features energy-absorbing features lifting and tie-down devices personnel barriers The staff verified that the drawings include the information described in NUREG-2216 on the (1) materials of construction, (2) dimensions and tolerances, (3) codes, standards, or other specifications for materials, fabrication, examination, and testing (4) welding specifications, including location and nondestructive examination (NDE), (5) coating specifications and other special material treatments that perform a safety function and (6) specifications and

24 requirements for alternative materials. The staff determined that the drawings for the package provide the necessary information identified in the NRC guidance documents and the engineering drawings provided by the applicant are consistent with the design and description of the package, in accordance with 10 CFR 71.33, Package Description. Therefore, the staff determined that the drawings provided by the applicant were acceptable.

2.2.2 Codes and standards The applicant identified the codes and standards, Federal regulations and NRC guidance documents (NRC Bulletins, Regulatory Guides, NUREG Documents) used for controlling the design and fabrication of components of shipping containers used for transporting radioactive materials in SAR section 1.1.5. The applicant stated that the criteria selected from the ASME B&PV Code and are based on the level of radioactive materials being transported and the nuclear safety function of the container's components. The applicant identified the following codes and standards for the materials, design, construction, nondestructive examination (NDE),

and testing for the TN Eagle transportation package:

ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsections NB, NC, NF, and NG, 2017.

ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsections NCA, 2017.

ASME Boiler and Pressure Vessel Code,Section II, Materials Specifications, Parts A, B, C and D, 2017.

ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendices, 2017.

ANSI N14.5, Leakage Tests on Packages for Shipment of Radioactive Materials, 2014.

Aluminum Standards and Data, Volume 1, The Aluminum Association, 1976.

The applicant stated in SAR section 7.2 that the containment boundary of the TN Eagle package is designed and constructed in accordance with the ASME B&PV Code, 2017 edition.

The applicant stated that the TN Eagle is designed and constructed without ASME B&PV Code Case usage. The applicant stated that the TN Eagle package containment boundary is designed and fabricated as a Class 1 component in accordance with the ASME B&PV Code,Section III, Division 1, Subsection NB. The applicant provided alternative provisions to the ASME B&PV Code in SAR chapter 1, section 1.6.8.

The applicant stated that the affected parts are identified as NB in the code criteria column of the drawing parts list. The applicant stated that attachments to the containment boundary with a pressure retaining function, or in the support load path for the packaging body, are identified as NF on the drawing parts list and stated that the stress analysis rules of ASME B&PV Code,Section III, Division 1, subsection NF are used for these parts. The applicant stated that the attachments to the containment boundary are constructed using either ASME or ASTM materials. The applicant stated that the parts that are not designated as ASME B&PV Code Section III, Division 1, Subsections NB or NF are designated as non-code on the drawing parts list. These include the impact limiters, neutron shielding, and non-pressure retaining cover plates. Materials specifications for these parts are included on the drawing parts list.

The applicant provided descriptions of the DSC designs that will be transported using the TN Eagle package in SAR chapter 1. These DSCs include the 37PTH and 89BTH DSCs approved in CoC No. 1042, the 32PT and 32PTH1 DSCs approved in CoC No. 1004, the 24PT1 and 24PT4 DSCs approved in CoC No. 1029, and the FO, FC, and FF DSCs approved in the SNM license 2510. The applicants description of the DSCs included drawings, material

25 specifications, identification of the ASME B&PV Codes used in the DSC design, construction, and inspection, and a list of ASME B&PV Code exceptions for each of these DSCs.

The staff reviewed the applicable codes and standards and material specifications for the TN Eagle packaging components. The staff determined that the use of ASME B&PV Code Section III, Division 1, Subsection NB for the containment boundary is consistent with the NRC guidance in NUREG-2216 which references NUREG/CR-3854, Fabrication Criteria for Shipping Containers. Table 4.1 of NUREG/CR-3854 provides guidance for use of ASME B&PV Code Section III, Division 1, criteria for the fabrication of containment, criticality, and other safety components for Category I containers such as those that transport SNF. Therefore, the staff finds these codes and standards and materials specifications to be acceptable.

The staff determined that the applicant has accurately identified the codes and standards used for the design and construction of the TN Eagle packaging and the DSCs that will be transported. The staff determined that the information provided by the applicant identifies the quality category or safety classification of the component and identifies the section of the ASME B&PV Code used in the design and construction. The staff verified that the confinement boundary of the TN Eagle packaging is designed and constructed using the criteria in ASME B&PV Code Section III, Division 1, Subsection NB, using either materials that are approved by ASME for the construction of Class 1 components or materials that the applicant has evaluated to meet the safety requirements for containment under NCT and HAC. The staff verified that the applicant identified non-ASME B&PV Code materials and has provided temperature dependent properties of these materials.

The staff determined that the applicant described and provided a basis for the code alternatives for the TN Eagle transportation packaging and the DSC contents in SAR section 1.6.8.

Therefore, the staff determined that the description of the codes and standards applicable to the TN Eagle package provided by the applicant was acceptable.

2.2.3 Weld design and inspection The applicant stated that all welding procedures and personnel shall be qualified in accordance with the ASME B&PV Code,Section IX. The applicant stated that weld inspections use the criteria of NB-5000 for containment boundary parts, and either NB-5000 or NF-5000 for all other components. The applicant described the TN Eagle transportation packaging body as a monobloc, or single piece forged carbon steel body with a bolted lid. The TN Eagle transportation packaging has wear-resistant rails welded to the interior surface of the packaging body and a weld overlay on the forged packaging body sealing flange. Weld inspection requirements for the rails and the weld overlay are included in the drawings provided by the applicant.

The applicant stated that the forged packaging body monobloc fabrication has no welding except for the stainless-steel overlay at the flange sealing surface, and welds to attach the rails to the inside surface. The applicant stated in SAR section 7.3.1 that this fabrication assures that the inleakage of water is not credible after HAC drop events. The applicant stated in SAR section 4.3 that the TN Eagle is designed and tested to be leak-tight per ANSI N14.5. The applicant further states in this section that the results of the structural and thermal analyses presented in chapters 2 and 3, respectively, demonstrate that the package will remain leak-tight and, thus, meet the leakage criteria of 10 CFR 71.51 and prevent ingress of water for all the HAC. Section 6.1 of the SAR further describes the design features credited to prevent water ingress due to immersion in accordance with 10 CFR 71.55(e). These statements, coupled with the conclusions provided by the review conducted in chapters 2 and 3 of this SER, provide a

26 basis that the package excludes water under HAC which the staff finds acceptable for demonstrating compliance with 10 CFR 71.55(e).

The impact limiters for the TN Eagle transportation package are constructed using an aluminum honeycomb material encased in a stainless-steel shell. The applicant provided weld inspection methods for the impact limiter shell welds in the TN Eagle package drawings. The applicant described the periodic inspection of the impact limiter casing welds in the SAR chapter 9, Acceptance Tests and Maintenance Program Evaluation.

The staff reviewed the drawings for the TN Eagle package and the weld design and inspection methods. The staff noted that the TN Eagle packaging body is a forged component with no welded structural components. The staff determined that welds and inspection methods for the rail attachments were acceptable because the applicant identified the weld locations, the components to be welded, the functions of the components, and the weld inspection methods.

The staff determined that the welds and inspection methods for the stainless-steel impact limiter casings were acceptable because the applicant identified the weld locations, materials, and inspection methods. The staffs review of the applicants periodic inspection of the impact limiters is included in this SER.

2.2.4 Mechanical properties The applicant provided a description of the mechanical properties of the packaging materials in SAR section 7.4 which included tensile properties, fracture resistance, and the properties of the impact limiter materials. The applicant stated that the TN Eagle package does not use aluminum parts other than the honeycomb of the impact limiters and the bottom spacer plate. The honeycomb is subject to neither high temperature nor tensile loads in NCT, and the spacer plate is replaceable dunnage; therefore, tensile properties, and creep of aluminum alloys at elevated temperatures, are not required for the evaluation of the TN Eagle package.

The applicant provided tables in chapter 7 of the SAR listing the mechanical properties for the containment boundary, shielding materials, impact limiter materials, and materials used for the structural components of the TN Eagle transportation packaging. The applicant included material properties including elastic modulus, yield strength, tensile strength allowable stress or allowable stress intensity as a function of temperature.

For ASME B&PV Code materials, the applicant cited the material property values included in the ASME B&PV Code,Section II, Part D and provided properties as a function of temperature. For ASTM or other non-ASME materials, the applicant provided supplemental information describing the testing methods and results used to determine the mechanical properties of the materials as a function of temperature. The applicant also provided values for Poissons ratio for all materials. For the containment component materials, the applicant also provided ductility values from tensile testing. The applicant also provided minimum toughness values or specified toughness testing requirements for containment component materials.

The staff reviewed the material properties provided by the applicant and determined that the applicant provided the mechanical properties of the materials used for the containment boundary, shielding, impact limiters and structural components of the TN Eagle transportation packaging. The staff reviewed ASME B&PV Code materials properties as a function of temperature and determined that that the properties provided by the applicant were acceptable because the applicant used the values provided in ASME B&PV Code Section II, Part D.

27 For containment materials that are not included in ASME B&PV Code Section II, Part D, the staff determined that the ASTM materials specified by the applicant were acceptable because these specifications have controls on composition, processing and material properties and the applicant supplied materials properties as a function of temperature that were supported by testing data. The staff determined that the temperature ranges for the mechanical properties provided by the applicant bound the range of the packaging component temperatures provided in chapter 3 of the SAR for NCT and HAC conditions.

Therefore, the staff determined that the mechanical properties of the materials for the TN Eagle transportation package provided by the applicant were acceptable.

2.2.5 Thermal properties of materials The applicant provided thermal properties of the materials including thermal expansion coefficients, thermal conductivity and heat capacity. The applicant provided values of the thermal properties as a function of temperature obtained from the ASME B&PV Code Section II Part D. In addition, for the containment components and the neutron shield ring the applicant provided additional information including emissivity. The applicant also provided emissivity and emissivity of coatings used on ferritic components. The staff determined that the thermal properties provided by the applicant were acceptable because they were supported by information included using the ASME B&PV Code Section II, Part D, and the specific testing to obtain the thermal properties of the materials and coatings.

2.2.6 Radiation shielding The staff reviewed information provided by the applicant regarding neutron shielding materials for the TN Eagle package. The applicant states that neutron shielding is provided by the proprietary VYAL-B resin blocks inserted in the shielding rings and in the top of the impact limiters. The applicant stated that the maximum temperature during NCT may slightly exceed the material temperature limit. The applicant asserts that the region of the resin that is above the long-term limit is very small compared to the overall mass of the resin. The region consists of a portion of the middle six resin blocks with only a small thickness and limited volume of the resin block exceeding the VYAL-B long term temperature limit.

The staff reviewed the SAR and the applicants response to an RAI on the thermal compatibility of the VYAL-B resin neutron shielding material. The staff determined that the applicants analysis of potential thermal effects on the VYAL-B resin shielding material was conservative and would result in a small increase in the dose rate at the package surface under NCT if the maximum long term temperature limit of the VYAL-B resin was exceeded as indicated in the applicants thermal analysis. Therefore, the staff determined that the thermal evaluation of the VYAL-B resin neutron shielding materials provided by the applicant was acceptable.

The SAR provides specification requirements for the composition of the VYAL-B resin and lists a minimum resin density and acceptance ranges for hydrogen and boron content. As these characteristics are part of the SAR, formulation of this resin is considered controlled for these specific characteristics and changes or substitutions would be subject to an amendment.

Polymers generally have a large coefficient of thermal expansion compared to metals. This can result in shrinkage of polymers under cold temperature conditions. The staff evaluated provisions in the neutron shield design to assure that excessive neutron streaming will not occur as a result of this shrinkage. The applicant stated that based on the configuration of the resin blocks, streaming in circumferential gaps would not be possible even with resin shrinkage, due to the blocks being staggered. Concerning the axial direction, if the external temperature and the DSC temperatures are extremely low, a small gap can form. The applicant states that the

28 low activity of the spent fuel at the low temperature and overall dose rate remains bounded by other cases. The staff reviewed the detailed drawing and material property information provided by the applicant and determined that differential thermal expansion of the TN Eagle package would not result in neutron streaming.

The staff reviewed the application information to determine the ability of the neutron shielding materials to withstand the combined aging effects of heat and radiation. The applicant stated that the VYAL-B resin composition used in the shielding analysis is the minimum guaranteed composition, which results in lower hydrogen and boron contents (-4 percent and -9 percent, respectively) and lower density (-2 percent) compared to the nominal composition. These lower values may account for some potential degradation that could occur over time due to the combined aging effects of heat and radiation.

The applicant performed a calculation including VYAL-B resin used, the neutron source per fuel assembly, total assemblies to calculate a conservative energy deposition for 10 years and found it to be below the threshold for damage due to long term exposure. The staff reviewed the calculation provided by the applicant for heat and radiation and found it to be conservative for evaluating the aging effects of heat and radiation.

The applicant states that gamma shielding is provided by the forged cask body, the primary lid, RACP, top handling ring, and the metallic parts of the shielding rings. The staff verified that the application describes the physical dimensions of the gamma shielding materials, namely that the dimensions of the components are provided in the drawings. The applicant indicates that acceptance testing for these materials is fulfilled by visual and dimensional inspections, as described in SAR section 9.1.1. The staff reviewed the acceptance tests in SAR section 9.1.1 and determined that the gamma shielding materials are fabricated to industry standard specifications that ensure uniform material properties. Therefore, the staff determined that the applicant has adequately identified material specifications and acceptance tests to ensure the composition and dimensions of the shielding materials.

The staff determined that the applicant has met the requirements of 10 CFR 71.43(f), 71.51(a),

and 71.64(a) and has adequately described the acceptance testing for gamma-and neutron-shielding materials, as described in NUREG/CR-3854.

2.2.7 Criticality control The applicant stated that the TN Eagle transportation packaging does not have criticality control materials. The applicant stated that the criticality control is addressed by materials that are part DSCs which are the contents of the TN Eagle package. The applicant stated that the neutron-absorbing (poison) material specifications, the percent credit for boron-based neutron absorbers and the qualifying properties not associated with neutron attenuation are addressed in the authorized content DSCs. The staffs review of the TN Eagle transportation packaging contents including the DSCs, spent fuel, and criticality control materials are included in this SER.

2.2.8 Corrosion resistance The applicant provided a description of operating environments for the TN Eagle in SAR section 7.7.4. In addition, the applicant described the coatings and treatments to protect the carbon steel components from corrosion. The applicant also described the effects of the operating environments on the stainless-steel components of the TN Eagle packaging. The staffs review of the information provided by the applicant are included in the following subsections:

29 Environment The applicant stated that the TN Eagle transportation package is designed for dry loading DSCs from and to the HSM, so there is no water immersion under normal operations. The applicant stated that the operating environment is limited to normal atmospheric exposures including rain and snow and possible exposure to road salts during transfer from the HSM to a rail car or during transport.

The applicant stated that the DSCs that form the TN Eagle transportation packaging contents are protected from direct exposure to precipitation and are exposed only to the humidity and aerosols in the cooling air that flows through the HSM during their storage period. The applicant stated that the internal environment of the DSCs is dry helium. The loading procedures included by the applicant in SAR section 8.1.2 state that the void space inside the TN Eagle transportation packaging and the DSC is backfilled with helium and as a result, the applicant stated that the internal environment is not corrosive to the DSC stainless steel surfaces.

The staff reviewed the description of the environments provided by the applicant in the SAR section 7.8 and the, the operating procedures in TN Eagle SAR chapter 8. The staff determined that the applicant has accurately described the range of operating environments for the TN Eagle transportation package, and therefore, the applicants description was acceptable. The staff evaluation of the corrosion resistance of the TN Eagle transportation packaging materials is included in this SER.

Carbon and low alloy steels The applicant stated that all exposed surfaces of carbon and low alloy steels are coated to prevent corrosion. The applicant described the coatings used protect the carbon and low alloy steel components in SAR section 7.8. The applicant stated that silicone sealant is applied between the shielding rings, top handling ring, bottom ring, bottom closure plate, and forged cask body to prevent water ingress to unpainted carbon steel surfaces.

The applicant stated in SAR section 9.2.3 that the coatings on the exterior surfaces of the cask components shall be inspected for damage or degradation within 12 months prior to releasing the loaded cask for shipment. The applicant stated that areas of the coating that show blistering, cracking, flaking, peeling, or other similar damage that can expose the base metal shall be evaluated and repaired as necessary.

The staff reviewed the description of the environments provided by the applicant in the SAR section 7.7.4, and the description of the coating for the corrosion protection of the carbon and low alloy steel surfaces. In addition, the staff reviewed the coating inspection requirements and inspection acceptance criteria described in the Acceptance Tests and Maintenance Program Evaluation in TN Eagle SAR chapter 9. The staff determined that the applicant s description of the corrosion protection methods and the coating inspection and acceptance criteria are sufficient to maintain the corrosion protection coatings for the carbon and low alloy steel materials. Therefore, the staff finds the applicants description to be acceptable.

Austenitic stainless steel The applicant stated that the only austenitic stainless parts of the TN Eagle transportation package containment are the weld overlay on the forged cask body top flange, the RACP port sealing surfaces, and the primary lid port plug. The applicant stated that the weld overlay will have little exposure to chloride aerosols because it is covered by the impact limiter during transport. The applicant stated that stress corrosion cracking (SCC) or pitting of the weld

30 overlay would not affect its design function performance unless a defect crossed the seal area; but SCC or pitting that damaged the seal area would be detected by pre-shipment and periodic leak testing. The applicant stated that the lid port plug is protected from the ambient environment by the lid orifice cover plate, and the lid port plug does not include welds, so it will not experience SCC. In addition, the applicant stated that the lid port plugs design function performance is also verified by leak testing. The applicant stated that the lid orifice cover plate and various not important to safety (NITS) spacers, pins, washers, DSC rails, etc., are also made of austenitic stainless steel.

The applicant stated that the primary lid, RACP, and the impact limiters structure are made of stainless steel, which is sufficiently corrosion-resistant for the environment described in section 7.7.5. The applicant stated that the top handling ring, the primary lid, and the ram access cover plate are SA-182 F6NM, which is a martensitic stainless steel. The applicant stated that these components do not contain welds and therefore, these parts are not susceptible to stress corrosion cracking.

The applicant stated that the shell of the impact limiters is welded stainless steel and could be subject to pitting or stress corrosion cracking due to environmental chlorides. The applicant stated that the shells will be visually inspected periodically, and indications of corrosion can be examined in more detail. Any defect that penetrates the shell can be found by periodic leak testing of the impact limiters.

The staff reviewed the stainless-steel components of the TN Eagle packaging, the corrosion assessment provided by the applicant and the Acceptance Tests and Maintenance Program Evaluation in TN Eagle SAR chapter 9. The staff determined that the applicants assessment of the corrosion resistance of the stainless-steel components was accurate and therefore acceptable. The staff determined that the applicants assessment that SCC or pitting corrosion of the weld overlay on the TN Eagle package containment components would be detected by inspections or leak testing was accurate and therefore acceptable. The staff determined that the applicants assessment of the SCC resistance of the martensitic stainless-steel components were accurate and therefore acceptable. The staff reviewed the periodic inspection and maintenance requirements for the impact limiters and confirmed that periodic leak testing is required every 5 years. The staff determined that the inspection requirement for the impact limiters was sufficient to detect damage and prevent corrosion related degradation. Therefore, the staff determined that the applicants assessment was acceptable.

2.2.9 Protective coatings Section 7.9 of the SAR provides a description of the protective coatings used in the TN Eagle transportation package. Additionally, chapter 1 mentions a corrosion protection coating on the bearing surfaces of the seals of the primary lid and RACP, but because this is later referred to as a weld overlay in section 7.8.3 of the application, the staff is not considering this a protective coating for the scope of this review area. Coatings considered in this section include:

Silicone-acrylic paint which will be used on the exposed carbon and low alloy steel on the outside surface of the cask body, Zinc-aluminum thermal spray on the inside.

The applicant provided data on the silicone-acrylic paint through reference to NTE-20-031053-000 Rev. 1, International system coating qualification, which provides a sufficient level of detail for the silicone-acrylic coating specification. The applicant indicates that manufacturer recommendations are followed for surface preparation, primer coat selection, and coating application. The staff notes that these details are also reflected in the NTE-20-031053-000 Rev.1 - International system coating qualification document referenced in the application.

31 The applicant provided data on the zinc-aluminum thermal spray through reference to RAP 00033156 Revision 0, Report for Emissivity Measurements of Metallic Thermal Spray Aged at Elevated Temperatures. The staff reviewed the report, which concludes that zinc-aluminum spray mechanical properties are unaffected after aging at 100°C (212°F) and 250°C (482°F).

Additionally, abrasion of the coating is not expected during normal operations and post aging at 100°C (212°F) and 250°C (482°F) the thermal spray resists scratching by nail.

The staff reviewed the SAR to note the scope of coating application. The SAR indicates that the silicone-acrylic paint will be used on the exposed carbon and low alloy steel on the outside of the cask body. Per 7.8.1 of the SAR, the applicable environment for the TN Eagle transportation package exposes it to rain, snow, and possible road salts. The staff verified that the coatings would not react with the package internal components and contents and would remain adherent and inert when the package is loaded, unloaded, or transported.

The staff reviewed the service environments and data for the coating systems and verified that the candidate coating systems for the steel and stainless-steel components are appropriate for the applications. The maintenance program described in section 9.2 of the SAR includes periodic visual inspections, including an assessment of coatings on the exterior surfaces of cask components for damage (e.g., blistering, cracking, flaking, peeling) or degradation and the performance of coating evaluations and repairs, if necessary.

The outer shell surface is credited with an emissivity of 0.86, which relies on the silicone-acrylic paint to ensure such a value. Page 3-30 of the SAR states: To account for uncertainties and to provide flexibility in selection of alternate coatings, the thermal analysis uses a solar absorptivity of 0.613 and an emissivity of 0.86 for the external painted surfaces of the TN Eagle-STC transportation package and impact limiter surfaces as stated in Chapter 7. If a different combination of coatings is considered, the emissivity and absorptivity will be verified by testing to ensure the values utilized in the thermal evaluations remain bounding. If they do not remain bounding, the thermal evaluation in Section 3.3 for NCT, Section 3.4 for HAC, Appendix 3.6.3 for HBU fuels, Appendix 3.6.4 for damaged and failed fuels, and Appendix 3.6.5 for TN Eagle SC with non-EOS DSCs will be evaluated to ensure all criteria identified in Section 3.2.3 are satisfied.

The staff notes that section 7.11 of the SAR mentions that organic materials that could be affected by gamma irradiation includes the neutron shielding resin, closure O-rings, and silicone-acrylic exterior coating. The section later discusses the maximum total gamma energy deposit rate for the resin and PTFE.

As a result, based on the capability of the coatings to protect the metallic surfaces from corrosion (or enhance decontamination) and the maintenance practices that ensure that coatings remain intact, the staff finds the coatings to be acceptable.

2.2.10 Content reactions The applicant stated that because the TN Eagle package will be dry loaded and the TN Eagle package cavity will be back filled with helium before each transport, there are no possible flammable or explosive reactions. Similarly, the applicant states that there are no content chemical reactions, outgassing, or corrosion reactions for the TN Eagle package. The applicant stated that the content DSCs were vacuum-dried and backfilled with helium before storage.

32 The staff reviewed the applicants assessment and determined that flammable or explosive reactions will not occur for the TN Eagle package because the system will be loaded under dry conditions and the TN Eagle packaging materials are compatible with the stainless-steel DSC under dry conditions. The staff determined that any residual moisture that may be present either inside the TN Eagle transportation packaging or on the surface of the DSC being loaded will be removed during the loading by the evacuation and helium backfilling of the TN Eagle packaging cavity as described in SAR section 8.1.2. The staff determined that the applicants assessment of no content chemical reactions, outgassing, or corrosion reactions for the DSCs transported as contents in the TN Eagle package is acceptable because drying and helium backfilling of the DSC prior to placement in storage removes water and created an inert environment inside the DSCs. Therefore, the staff determined that the applicants assessment that there will be no content reactions was acceptable.

2.2.11 Radiation effects The applicant stated that at the center of the content DSCs, the neutron fluence is on the order of 1015 n/cm2 after 20 years. Further the applicant stated that the fluence experienced by the metal components of the TN Eagle transportation packaging will be much lower because the transportation packaging which contains the DSCs is exposed to a lower neutron fluence for a limited time during transportation. The applicant stated that because neutron embrittlement is initiated around 1017 n/cm2, there will be no effect of neutron irradiation on the metal components of the TN Eagle transportation packaging. The applicant stated that metals are not affected by gamma irradiation.

The applicant stated that organic materials that could be affected by gamma irradiation are the neutron shielding resin, the closure O-rings, and the silicone-acrylic exterior coating. The applicant stated that the maximum total gamma energy deposit rates are 7.33 105 rad/s for the resin and 6.65 105 rad/s for the fluorocarbon components of the seals. The applicant stated that the maximum gamma exposures in 40 years are 9.25 104 rad for the resin and 8.40 104 rad using a conservative assumption that the PTFE is located at the cask side with much higher gamma flux, instead of the actual locations at the cask ends.

The staff reviewed the applicants analysis using the guidance in NUREG-2216 Section 7.4.11.

The staff also noted that the TN Eagle packaging is limited to 800 one-way trips as stated in SAR section 9.2.3. The staff determined that neutron embrittlement of the metal components of the TN Eagle packaging components will not occur over the expected period of use.

In addition, the staff determined that the gamma radiation exposure of the neutron shielding resin is insufficient to result in radiation damage. The staff determined that gamma irradiation damage to the fluorocarbon components of the seals were unlikely as noted by the conservative analysis provided by the applicant. In addition, the staff determined that the inspection and maintenance program include inspection of the seals as stated by the applicant in SAR section 9.2.3. The staff determined that any radiation damage to the seals that affected seal performance would be identified in the inspection and maintenance program and damaged seals would be replaced. Therefore, the staff determined that the analysis provided by the applicant was acceptable and radiation damage of the cask components would not occur over the expected period of use.

2.2.12 Package Contents

33 The applicant included a description of the DSCs authorized as content for the TN Eagle transportation package in SAR section 1.2.3. These DSCs included the EOS-37PTH and the EOS-89 BTH from the CoC No. 1042 system, the 24PT1 and the 24PT4 from the CoC No. 1029 system, the 32 PT and 32 PTH1 from the CoC No. 1004 system and the FO, fuel with control components (FC), and FF DSCs from SNM-2510. For each DSC the applicant identified the components that make up the confinement boundary, described the SNF basket structure and identified the fuel assemblies that were authorized for storage.

In SAR section 1.5 the applicant provided the drawings for the DSCs that will be transported as contents in the TN Eagle package. The drawing packages for each DSC included the DDC confinement barrier with the structural and shielding components of the DSC. The applicant also provided drawings for the DSC basket that maintains the geometry of the SNF. The applicant also provided drawings for the damaged fuel basket cell endcaps and failed fuel cans. The drawings provided by the applicant included a bill of materials with material specifications, identification of applicable ASME B&PV Code design criteria, and quality category for ITS components. The drawings provided by the applicant also identified component dimensions, dimensional tolerances, welds and NDE requirements for the DSCs and internals components.

The applicant included tables for each DSC in SAR section 1.6.2 that describe the physical characteristics of the DSC such as size, weight, and materials of construction. The applicant also included descriptions of the neutron poison materials including the minimum boron-10 concentration for the neutron absorber materials in the DSC baskets for criticality control. The applicant provided the maximum decay heat per DSC, minimum fuel enrichment and minimum assembly cooling time. The applicant provided a description of the SNF authorized to be loaded into the DSCs which included the type of assembly, allowed control components, number of assemblies, maximum uranium loading per assembly, maximum assembly weight, maximum fuel burnup, fuel condition and, if permitted, the number of damaged and failed fuel assemblies The staff reviewed the information provided by the applicant to describe the package contents including the description of the TN Eagle package contents including the DSCs provided in SAR section 1.2.3, the drawings of the DSCs, and the DSC internal components in SAR section 1.5 and the technical information associated with each DSC provided in tables included in SAR section 1.6. The staff determined that the applicant provided a description of the TN Eagle package contents in SAR chapter 1 including the description of the DSCs in SAR section 1.2.3, the drawings in SAR section 1.5 and the technical information on the DSCs and the SNF contents in SAR section 1.6.

The staff determined that the applicant provided an acceptable description of the contents of the TN Eagle transportation package. The staff determined that the applicant provided information with respect to the drawings for the DSC contents because the applicant provided information that was in accordance with the guidance in NUREG/CR-6407 and NUREG/CR-5502. The staff determined that the technical information on the DSC contents provided in SAR section 1.6 was acceptable because the applicant identified the physical characteristics of the DSC, descriptions of the neutron poison materials and the descriptions of the SNF approved for the contents of the DSCs. Therefore, the staff determined that the information provided by the applicant was acceptable.

2.2.13 Bolting material The staff reviewed the information provided by the applicant pertinent to bolting material in section 7.14 of the application. The applicant provided information in the material tables of chapter 7 consistent with ASME B&PV Code Section II, Part D for material properties for the

34 bolting material using ASME materials. The applicant added for table 7-8, the bolts utilize tensile and yield stress values below the ASME Code values for the purposes of structural evaluation.

For those bolts not using ASME material, the staff noted that the property values are bounded by the ASME SA-540/540M B24/B23 Class 2 property values at elevated temperatures. The staff finds the bolting material acceptable in meeting 10 CFR 71.31(c)

The staff confirmed that the applicant identified the materials to be used in bolted connections in accordance with 10 CFR 71.33(a)(5). The applicant indicates that alloy steel bolts are coated with bi-chromatic zinc, which the drawings further reinforce with a note that states non-stainless threaded fasteners may have corrosion resistant coatings approved by Orano TN.

The staff observes that zinc coatings are commonly used to provide corrosion resistance. The applicant has specified bolting to be made of ASME SA-320 L43, ASME SA-520/ 540M B24/B23 Class 2, or Carbon Steel meeting the requirements of the Deutsches Institut für Normung (DIN) (i.e., German institute for standardization) 1.6580. The applicants drawing also indicates that a threaded insert option will be available for items V1, V3, V5, V10, V14, and V17 made of SAE AS7245 or equivalent. It is not required to be ASME Code material. The staff considers these material selections to be acceptable for corrosion resistance, considering the zinc coatings that will be applied to alloy steel bolts. Further, periodic inspections of the bolts and mating threaded holes are described in section 9.2.3 of the SAR as part of the maintenance program, which will allow for identification of damage or degradation and allow for rework or replacement prior to use. As such, the staff considers that the applicant has assessed the effects of corrosion, chemical reactions, and radiation effects on the bolting materials, in accordance with 10 CFR 71.43(d).

The staff reviewed the material tables in chapter 7 for the bolting materials specified in the SAR.

The tables provided acceptance criteria for mils of lateral expansion consistent with NB-2333-1, which ensure appropriate resistance to brittle fracture for the materials specified.

2.2.14 Seals The applicant stated that the primary lid and ram access cover plate are both equipped with two elastomeric seals and the cask body sealing surfaces are protected with a stainless-steel cladding or a corrosion protection coating as described in chapter 1. The applicant stated that other non-containment elastomeric seals are at the lid alignment pin, test ports, and lid port cover plate. The applicant stated in SAR section 7.7.4 that the main containment boundary seals are elastomeric, not metallic, and therefore not subject to corrosion. The applicant stated that the port seal is the only metallic containment boundary seal, and it is protected from the environment by the port cover and another seal which is also metallic but not a containment boundary seal. The applicant stated that both metallic seals will be replaced prior to each transport, eliminating corrosion concerns. The metallic seals have a minimum and maximum temperature rating of 40°C (40°F) and 340°C (644°F) respectively. The maximum metallic seal temperatures under HAC are 217°C (422°F) for intact fuel and 224°C (435°F) for HBU fuel, which are bounded by the maximum temperature rating.

The applicant states that the fluorocarbon seals have a temperature rating of 40°C (40°F) to 204°C (400°F), which bounds the expected maximum seal temperature of 169°C (336°F) under HAC for intact fuel and 179°C (354°F) for HBU fuel. The applicant uses the maximum temperature of the primary lid seal for the maximum fluorocarbon seal temperature, which they consider as conservative. The staff verified that the fluorocarbon seals reported maximum temperatures do not exceed the stated maximum temperature limit of 204°C (400°F) as

35 reported by the applicant. The staff further confirmed that the minimum operating temperature limit specified for the seals is 40°C (40°F), which is in compliance with 10 CFR 71.71 (c)(2).

The maintenance program specifies replacement of elastomeric seals within 12 months prior to releasing the loaded cask for shipment. The staff considers this replacement interval sufficient to mitigate any thermal-and radiation-induced aging to the seals.

2.2.15 Conclusion The NRC staff concludes, based on review of the statements, and representations in the application, that the materials used in the TN-Eagle transportation package design have been adequately described and evaluated and that the package meets the requirements of 10 CFR Part 71.

3.0 THERMAL EVALUATION 3.1 Review Objectives The objective of the review of TN Eagle spent fuel transportation package is to verify that the thermal performance of the package has been adequately evaluated for the tests specified under NCT and HAC and that the package design satisfies the thermal requirements of 10 CFR Part 71. This case is also reviewed to determine whether the package fulfills the acceptance criteria listed in section 3 of NUREG-2216, "Standard Review Plan for Transportation Packages for Spent Nuclear Fuel and Radioactive Material.

3.2 Description of the Thermal Design 3.2.1 Design Features The TN Eagle spent nuclear fuel transportation package consists of a forged body which conducts the decay heat to the cask outer surface. The other thermal design feature of the cask is the conduction path created by the shielding rings that contain the neutron shielding material.

The neutron shielding resin blocks are placed in the lodgments of the shielding rings. The shielding rings are designed to shrink fit tightly against the steel shell surfaces, thus improving the heat transfer across the neutron shield. Heat dissipates from the packaging outer surfaces via natural convection and radiation. The outer surfaces of the shielding rings are painted white to enhance the thermal radiation exchange with ambient. TN Eagle is equipped with impact limiters that provide protection to the lid and bottom regions from the external heat input due to fire during the HAC thermal event. A personnel barrier prevents access to the outer surfaces of the cask body. The barrier, which consists of a stainless-steel mesh attached to stainless steel tubing, encloses the cask body between the impact limiters.

The TN Eagle can accommodate the following dry shielded canister (DSC) types, as presented below:

36 Table: Maximum Heat load per DSC DSC Type TN Eagle-STC Cask Type Maximum DSC Heat Load for Transport (kW)

EOS-37PTH LC 38.40 EOS-89BTH LC 31.15 32PT SC 24.00 32PTH1 Type 1 SC 26.00 32PTH1 Type 2 SC 24.00 24PT4 SC 24.00 24PT1 SC 14.00 FO/FC/FF SC 13.50 The EOS-37PTH DSC consists of a shell assembly, which provides both confinement and shielding, and an internal basket which locates and supports up to 37 pressurized water reactor (PWR) spent nuclear fuel assemblies (FAs). The EOS-37PTH DSC basket structure consists of interlocking slotted plates to form an egg-crate type structure. The egg-crate structure forms a grid of 37 fuel compartments that house the PWR spent fuel assemblies (SFA). The egg-crate grid structure is composed of one or more of the following: a steel plate, an aluminum plate, and a neutron absorber (poison) plate. The aluminum plates, together with the poison plates, provide a heat conduction path from the SFA to the DSC rails and shell.

The EOS-89BTH DSC consists of a shell assembly, which provides both confinement and shielding, and an internal basket which locates and supports up to 89 boiling water reactor (BWR) spent nuclear FAs. The EOS-89BTH DSC basket structure consists of interlocking slotted plates to form an egg-crate-type structure. The egg-crate structure forms a grid of 89 fuel compartments that house the BWR SFA. The egg-crate grid structure is composed of one or more of the following: a steel plate, an aluminum plate, and a neutron absorber (poison) plate.

The aluminum plates, together with the poison plates, provide a heat conduction path from the SFAs to the DSC rails and shell.

Other non-EOS DSCs including 24PT4, 32PT, 32PTH1, FO/FC/FF, and 24PT1 DSCs can be transported in the TN Eagle SC with internal sleeves. The design features of 32PT and 32PTH1 DSCs are discussed in sections M.1.2.1 and U.1.2.1 of Updated Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, Revision 18, Docket No. 07201004, respectively. The design features of 24PT1 and 24PT4 DSCs are discussed in sections 1.2.1.1 and A.1.2.1.1 of Updated Final Safety Analysis Report for the Standardized Advanced NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, Revision 9, Docket No. 07201029, respectively. The design features of FO/FC/FF DSCs are discussed in section 1.2.2 of Volume I of Rancho Seco Independent Spent Fuel Storage Installation Final Safety Analysis Report, Revision 6, Docket No.

07200011. Design features for non-EOS DSC were reviewed and approved previously. The detailed descriptions of these DSCs are documented in SAR appendix 1.6.3 through appendix 1.6.7.

3.2.2 Thermal Design Criteria

37 Several thermal design criteria are established by the applicant for the TN Eagle package to ensure that the package meets all its functional and safety requirements. These criteria are listed below:

Maximum fuel cladding temperature limits of 752°F (400°C) for NCT and 1058°F (570°C) for HAC are considered for the FAs with an inert cover gas.

The maximum fuel cladding temperature limit of 752°F (400°C) is considered for high burnup (HBU) FAs under NCT.

However, HBU reconfigured fuels, whose physical integrity may not be guaranteed under HAC, are conservatively assumed as rubble. Therefore, the above maximum fuel cladding temperature limit of 1058°F (570°C) for HAC is not applicable to the HBU reconfigured fuels.

Containment of radioactive material and gases is a major design requirement. Seal temperatures must be maintained within specified limits (as specified in the application) to satisfy the leak tight containment requirement.

To maintain the stability of the neutron shield resin, a maximum allowable temperature of 320°F (160°C) is considered for the neutron shield for NCT.

To prevent melting of the gamma shield (lead) under NCT, an allowable maximum temperature of 621°F (327.5°C - melting point of lead) is considered for the gamma shield.

A temperature limit of 320°F (160°C) is considered for wood to prevent excessive reduction in structural properties at elevated temperatures.

The recommended temperature design limit for the aluminum honeycomb in the impact limiter is 248°F (120°C) under NCT, as specified in the application.

In accordance with 10 CFR 71.43(g) the maximum temperature of the accessible packaging surfaces in the shade is limited to 185°F (85°C).

The NCT ambient temperature range is -20°F to 100°F (-29°C to 38°C) per 10 CFR 71.71(b).

3.2.3 Contents Decay Heat The thermal analysis for the TN Eagle loaded with the DSCs listed in the table in SAR section 3.1.1.1 is based on a range of maximum total heat load of 13.5 kW to 38.4 kW per DSC. Table 1 above provides the maximum heat load per DSC type.

The permitted heat load zone configurations (HLZCs) for all DSCs are listed in Final Safety Analysis Report (FSAR) chapter 1, appendix 1.6.1 through appendix 1.6.7. These design basis HLZCs are symmetrical and show maximum allowable heat load per FA and per DSC, which result in bounding maximum fuel cladding and DSC component temperatures.

The staff reviewed the design features, design criteria, and contents decay heat of the TN Eagle spent nuclear fuel transportation cask. Based on the information provided in the application regarding these items, the staff determines that the application is consistent with guidance provided in section 3.4.1 (Description of the Thermal Design) of NUREG-2216.

38 Therefore, the staff concludes that the description of the thermal design is acceptable because the description satisfies NUREG-2216 and subsequently meets the requirements of 10 CFR Part 71.

3.2.4 Summary Tables of Temperatures The summary tables of key package component temperatures for NCT (SAR tables 3-8 and 3-

10) were reviewed. The temperatures are consistently presented throughout the SAR for NCT.

For the hypothetical accident conditions, the applicant presented the pre-fire, during-fire, and post-fire component temperatures (SAR tables 3-16 and 3-19). Except for the impact limiters resin (which is assumed to be decomposed or charred after the fire accident), all components remain below their material property limits (specified in SAR section 3.2). The temperatures and design temperature limits for the package components were reviewed and found to be consistent throughout the SAR.

3.2.5 Summary Tables of Maximum Pressures The bounding maximum internal pressures inside the TN Eagle cask cavity are calculated in SAR section 3.3.3 for NCT and section 3.4.4 for HAC. The bounding maximum internal pressures of the TN Eagle cask cavity are summarized in SAR table 3-12. These pressures remain below the design pressures for NCT and HAC considered for the structural evaluation.

The summary tables of the containment pressure under normal conditions of transport and hypothetical accident conditions were reviewed and found consistent with the pressures presented in the General Information, Structural Evaluation, and Containment Evaluation SAR sections. These tables reported the MNOP for normal conditions of transport and hypothetical accident conditions (fire). These pressures remain below the design pressures for NCT and HAC.

The staff reviewed the design description of TN Eagle transportation package thermal design and finds it acceptable. The staff reviewed the temperature and pressure design limits and calculated temperatures and pressures for the package and found them to be acceptable and consistent in the SAR.

3.3 Material Properties and Component Specifications 3.3.1 Material Properties The package application provided material thermal properties such as thermal conductivity, density, specific heat, and emissivity for all modeled components of the cask. The staff found these properties acceptable. The applicant specifies the natural convection heat transfer coefficient as a function of the product of Grashof and Prandtl numbers. This product is a function of length scale, surface-to-ambient temperature difference, and air properties. The thermal properties used for the analysis of the package were appropriate for the materials specified and for the conditions of the cask required by 10 CFR Part 71 during normal and accident conditions.

The staff reviewed the thermal properties used for the package analyses and determined that they were appropriate for the materials specified and for the package conditions required by 10 CFR Part 71 during NCT and HAC.

3.3.2 Component Specifications

39 The application provided component thermal technical specifications for the TN Eagle containment seals and poison plates used in DSC baskets. The applicant provided temperature ranges for the different materials. These ranges assure that the cask can be operated safety provided the thermal specifications for these materials are not exceeded. Maximum cavity internal design pressures for each DSC are provided in the application as well.

The staff reviewed the thermal properties used for the package analyses and determined that they were appropriate for the materials specified and for the package conditions required by 10 CFR Part 71 during NCT and HAC. The staff reviewed the component specifications for the TN Eagle and determined that the specifications were sufficiently clear to be evaluated as part of the thermal evaluation results.

3.4 Thermal Evaluation under Normal Conditions of Transport 3.4.1 Thermal Models The applicant developed separate TN Eagle thermal models for both EOS DSCs (37PTH and 89BTH). The thermal model includes the heat conduction and radiation within the DSC basket, impact limiters and TN Eagle cask; solar insolation through the TN Eagle cask and impact limiter external surfaces; and heat dissipation from the TN Eagle and impact limiter external surfaces via convection and radiation to the ambient. No convection is considered in the thermal models. The exterior surfaces of the TN Eagle cask, top and bottom impact limiters dissipate heat to the ambient via radiation and free convection and absorb heat from solar insolation during normal hot condition of transport. The applicant used these models to perform steady state evaluations for NCT conditions.

The staff reviewed the applicants description the TN Eagle transportation cask thermal models.

Based on the information provided in the application regarding the developed thermal models, the staff determines that the application is consistent with guidance provided in section 3.4.5 (General Considerations for Thermal Evaluations) of NUREG-2216. Therefore, the staff concludes that the description of the thermal models is acceptable because the description satisfies NUREG-2216 and subsequently meets the requirements of 10 CFR Part 71.

3.4.2 Heat and Cold The applicant performed steady state analysis using the TN Eagle LC thermal models without insolation to determine the accessible surface temperature of the impact limiters in the shade.

A heat load of 38.4 (37PTH DSC) and 31.15 (89PTH DSC) kW and boundary conditions at 100°F and no insolation are considered in the cask model to bound the maximum accessible surface temperature under shade. The comparison provided in SAR table 3-11 shows that all average temperatures determined for the EOS-37PTH DSC loaded in the TN Eagle LC cask are bounding for those of the EOS-89BTH DSC loaded in the TN Eagle LC cask. Based on this comparison, the applicant concluded that thermal evaluation performed for the TN Eagle LC loaded with the EOS-37PTH DSC with the maximum heat load of 38.4 kW remains bounding.

Therefore, no further evaluation is performed for the EOS-89BTH DSC loaded in the TN Eagle LC. SAR table 3-8 provides the bounding maximum temperatures of key components in TN Eagle LC loaded with EOS-37PTH DSC under NCT. All temperatures remain below their respective design limit.

The applicant calculated maximum accessible surface temperatures for impact limiter and personnel barrier. These temperatures are 149°F and 161°F, respectively. These temperatures are well below the limit of 185°F. In accordance with 10 CFR 71.43(g) the maximum

40 temperature of the accessible packaging surfaces in the shade is limited to 185°F (85°C).

Therefore, this regulatory requirement is met.

The applicant stated that under the minimum ambient temperature of -40°F (-40°C), the resulting packaging component temperatures will approach -40°F if no credit is taken for the decay heat load. Since the package materials, including containment structures and the seals, continue to function at this temperature, the minimum temperature condition has no adverse effect on the performance of the TN Eagle cask.

3.4.3 Maximum Normal Operating Pressure TN Eagle Cask Operating Pressure The applicant calculated the maximum pressures in the cask cavity for the TN Eagle LC loaded with EOS-37PTH and EOS-89BTH DSCs. The calculated pressures are presented in SAR table 3-12. The EOS-37PTH DSC in the TN Eagle LC with 38.4 kW heat load is bounding for the maximum TN Eagle LC cavity pressure for all DSCs.

Maximum Internal Pressure in EOS-37PTH DSC The applicant stated that SAR table 3-13 shows the bounding average temperatures of FAs and helium in DSC cavity determined for the EOS-37PTH DSC in the TN Eagle LC are lower than the temperatures determined for the EOS-37PTH DSC in storage and transfer. Therefore, the applicant concluded that the maximum internal pressures in table 4-45 of NUHOMS EOS System Updated Final Safety Analysis Report, remain bounding for the TN Eagle LC with the EOS-37PTH DSC during NCT and HAC of transport for intact, damaged, failed, and HBU FAs.

The MNOPs are below the containment design pressure, as reported in the application and therefore is acceptable. The staff reviewed calculations and results of the TN Eagle cask for NCT conditions and found them acceptable.

The staff reviewed the applicants analysis of the TN Eagle transportation cask during NCT.

Based on the information provided in the application regarding NCT analysis, the staff determines that the application is consistent with guidance provided in section 3.4.5 (Thermal Evaluation Under Normal Conditions of Transport) of NUREG-2216. Therefore, the staff concludes that the HAC analysis is acceptable because the analysis and results satisfy NUREG-2216 and subsequently meets the requirements of 10 CFR Part 71.

3.5 Thermal Evaluation under Hypothetical Accident Conditions The thermal model of the TN Eagle LC cask developed by the applicant to perform NCT is modified by the applicant in this evaluation to determine the maximum component temperatures for HAC. The applicant performed transient runs considering HAC conditions, per 10 CFR 71.71(c)(4).

3.5.1 Initial Conditions The initial temperatures for the TN Eagle LC cask transient model before the fire accident are determined using the same boundary conditions for NCT (100°F ambient with insolation) described in the SAR.

3.5.2 Fire Test Conditions

41 Based on the requirements in 10 CFR 71.73, a fire temperature of 1475°F, fire emissivity of 0.9 and a period of 30 minutes are considered for the fire conditions in the applicants thermal model. A bounding forced convection coefficient of 4.5 Btu/hr-ft2-°F is considered during burning period based on data from the report entitled Thermal Measurements in a Series of Long Pool Fires, SANDIA Report, SAND 85-0196, TTC-0659, 1987. Surface emissivity of 0.8 is considered for the packaging surfaces exposed to fire, based on 10 CFR 71.73.

3.5.3 Maximum Temperatures and Pressure The maximum component temperatures for transient runs are listed in table 3-16 of the SAR.

The seals are not explicitly considered in the models. This table shows that the maximum temperatures of the TN Eagle LC components calculated for HAC are lower than the allowable limits. The maximum seal temperatures are retrieved from the models by selecting the nodes at the locations of the corresponding seals. Most of the maximum temperatures determined for the EOS-37PTH DSC in the TN Eagle LC at 38.4 kW under HAC will bound the maximum temperatures for EOS-89BTH DSC in the TN Eagle LC at 31.15 kW.

The maximum internal pressures inside the cask cavity and inside the 37PTH and 69BTH DSCs are summarized in tables 3-12 and 3-13 of the SAR. These pressures are below the design limits specified in the SAR.

3.5.4 Maximum Thermal Stresses Thermal stresses for the TN Eagle cask are discussed in chapter 2 of the SAR.

3.5.5 Accident Conditions for Fissile Material Packages for Air Transport The TN Eagle package is not designed for air transport.

The staff reviewed the applicants analysis of the TN Eagle transportation cask during HAC.

Based on the information provided in the application regarding HAC analysis, the staff determines that the application is consistent with guidance provided in section 3.4.6 (Thermal Evaluation under Hypothetical Accident Conditions) of NUREG-2216. Therefore, the staff concludes that the HAC analysis is acceptable because the analysis and results satisfy NUREG-2216 and subsequently meets the requirements of 10 CFR Part 71.

3.6 Mesh Sensitivity The applicant performed a grid convergence study to obtain the discretization error for the bounding transport configuration. The discretization error is determined using the five steps specified in section 2-4.1 of ASME Verification and Validation (V&V) 20-2009 (American Society of Mechanical Engineers, Standard for Verification and Validation in Computational Fluid Dynamics and Heat Transfer, ASME V&V 20-2009, November 30, 2009).

3.7 Thermal Evaluation of High Burnup Fuel Assemblies The applicant stated that for the case that the physical configuration of the FAs is not altered (for both low burnup and HBU FAs), the thermal evaluations presented for NCT in SAR section 3.3 (for NCT) and in SAR section 3.4 (for HAC) remain valid for HBU FAs.

42 The applicant performed additional the thermal evaluations for NCT and HAC for the case that physical configuration of the FAs is not guaranteed to provide assurance that the containment of the TN Eagle cask is maintained during transportation of high burnup fuel. The evaluations provided in SAR section 3.6.3 fire conditions. SAR tables 3-8 and table 3-9 provide the bounding maximum and average temperatures for NCT and tables 3.6.3-4 and 3.6.3-7 provide the bounding maximum and average temperatures for HAC during transportation of HBU FAs in the TN Eagle LC loaded with the EOS-37PTH DSC.

The staff reviewed the applicants analyses of the TN Eagle transportation cask during NCT and HAC for altered physical configuration of FAs. Based on the information provided in the application regarding these analyses, the staff determines that the application is consistent with guidance provided in section 3.4.5 (Thermal Evaluation under Normal Conditions of Transport) and section 3.4.6 (Thermal Evaluation under Hypothetical Accident Conditions) of NUREG-2216. Therefore, the staff concludes that the HAC analysis is acceptable because the analysis and results satisfy NUREG-2216 and subsequently meets the requirements of 10 CFR Part 71.

3.8 Thermal Evaluation of Damaged and Failed Fuel Assemblies in TN Eagle The applicant evaluated the thermal performance of the TN Eagle LC loaded with the EOS-37PTH DSC with damaged or failed FAs under normal conditions of transport (NCT) and hypothetical accident condition (HAC). The evaluations provided in SAR section 3.6.3 provide bounding thermal results that satisfies the design criteria specified in SAR section 3.2.3 and the containment criteria of the TN Eagle LC with the EOS-37PTH DSC with damaged and failed FAs.

The staff reviewed the applicants analyses of the TN Eagle transportation cask during NCT and HAC for altered physical configuration of FAs. Based on the information provided in the application regarding these analyses, the staff determines that the application is consistent with guidance provided in section 3.4.5 (Thermal Evaluation under Normal Conditions of Transport) and section 3.4.6 (Thermal Evaluation under Hypothetical Accident Conditions) of NUREG-2216. Therefore, the staff concludes that the HAC analysis is acceptable because the analysis and results satisfy NUREG-2216 and subsequently meets the requirements of 10 CFR Part 71.

3.9 Thermal Evaluation of TN Eagle SC with Non-EOS DSCs In SAR section 3.6.5 the applicant described the approach to demonstrate that the evaluations and results provided for NCT and HAC for the TN Eagle cask loaded with EOS-37PTH canister bound the non-EOS DSCs included in table 1. The approach as described in the SAR consists of comparing the DSC shell temperature profiles obtained for storage and transfer conditions with those evaluated for transport. The applicant demonstrated in the SAR that the DSC shell temperatures obtained for storage conditions bound those obtained for transport in the TN Eagle cask. Therefore, the applicant concluded that DSC component temperatures (including maximum fuel cladding temperature) would also bound those for transport conditions. The same reasoning was applied for obtaining the maximum pressures. Therefore, the applicant concluded that no additional analyses are required for non-EOS DSCs.

The staff reviewed the applicants analyses of the TN Eagle transportation cask with non-EOS DSCs during NCT and HAC. Based on the information provided in the application regarding these analyses, the staff determines that the application is consistent with guidance provided in section 3.4.5 (Thermal Evaluation under Normal Conditions of Transport) and section 3.4.6 (Thermal Evaluation under Hypothetical Accident Conditions) of NUREG-2216. Therefore, the

43 staff concludes that the analyses are acceptable because the analyses and results satisfy NUREG-2216 and subsequently meet the requirements of 10 CFR Part 71.

3.10 Thermal Tests A thermal test of the TN Eagle fabricated cask is performed as described in SAR section 8.1.8 of the SAR. This thermal test if performed to measure the effective thermal conductivity of the cask in the radial direction. An analysis is to be performed to compare the thermal performance of the cask with the predicted analysis results. This is only required on the first cask (any model) fabricated to the design provided in the applicable drawings for package approval.

The applicant stated in the SAR that a periodic thermal test is not needed because degradation of materials is not expected to impact package thermal performance. The staff finds the explanation acceptable because the analysis provided bounding results.

3.11 Confirmatory Analyses The staff reviewed the thermal models developed by the applicant to perform the thermal evaluation of the TN Eagle transport package. The staff checked the code input in the calculation packages and confirmed that the proper material properties and boundary conditions were applied. The engineering drawings were also consulted to verify that proper geometry dimensions were translated to the analysis model. The material properties presented in the SAR were reviewed to verify that they were appropriately referenced and used.

3.12 Findings The staff reviewed the package description and evaluation and found reasonable assurance that they satisfy the thermal requirements of 10 CFR Part 71.

The staff reviewed the material properties and component specifications used in the thermal evaluation and found reasonable assurance that they are sufficient to provide a basis for evaluation of the package against the thermal requirements of 10 CFR Part 71.

The staff reviewed the methods used in the thermal evaluation and found reasonable assurance that they are described in sufficient detail to permit an independent review, with confirmatory calculations, of the package thermal design.

The staff reviewed the accessible surface temperatures of the package as it will be prepared for shipment and found reasonable assurance that they satisfy 10 CFR 71.43(g) for packages transported by exclusive-use vehicle.

The staff reviewed the package design, construction, and preparations for shipment and found reasonable assurance that the package material and component temperatures will not extend beyond the specified allowable limits during NCT, consistent with the tests specified in 10 CFR 71.71.

The staff reviewed the package design, construction, and preparations for shipment and found reasonable assurance that the package material and component temperatures will not exceed the specified allowable short-time limits during hypothetical accident conditions, consistent with the tests specified in 10 CFR Part 71.

44 Based on review of the statements and representations in the application, the staff concludes that the impact of the proposed changes on the thermal design has been adequately described and evaluated, and that the thermal performance of the TN-EAGLE package meets the thermal requirements of 10 CFR Part 71.

4.0 CONTAINMENT EVALUATION The objective of the review is to verify that the containment performance of the Model No. TN Eagle package has been adequately evaluated for the tests specified under both NCT and HAC of transport and that the package design satisfies the containment requirements of 10 CFR Part 71.

4.2 Description of Containment System The containment boundary of the TN Eagle package (shown in Figure 4-1 of the application) consists of a thick forged body, a primary lid, a lid port plug, a ram access cover plate, and the associated closure seals. The containment vessel is assembled with bolted connections and there are no associated welds. The containment vessel prevents leakage of radioactive material from the cask cavity. It also maintains an inert atmosphere (helium) in the cask cavity. The entire containment vessel, including each penetration is designed to meet the leak-tight criteria, per ANSI N14.5-2014. The lid port plug seal, and the inner seals in the primary lid and ram access cover plate, are the primary containment boundary seals. Outer seals are provided in the primary lid and ram access cover plate to facilitate leak testing of the inner containment seals.

4.3 Containment under Normal Conditions of Transport The entire containment vessel, including each penetration is designed to meet the leak-tight criteria, per ANSI N14.5-2014. The structural and thermal analyses presented in chapters 2 and 3 of the application, demonstrate that the cask remains leak-tight under any of the normal conditions of transport, which ensures there will be no release of radioactive material or ingress of water during transportation.

4.4 Containment under Hypothetical Accident Conditions The entire containment vessel, including each penetration is designed to meet the leak-tight criteria, per ANSI N14.5-2014. The results of the structural and thermal analyses presented in chapters 2 and 3, respectively, demonstrate that the package will remain leak-tight and, thus, meet the leakage criteria of 10 CFR 71.51 and prevent ingress of water for all the HAC.

4.5 Leakage Rate Tests The application states that the TN Eagle leakage testing is performed in accordance with the requirements of ANSI N14.5-2014. The acceptance criterion for the fabrication, periodic, maintenance, and pre-shipment leak testing is specified in the application.

4.6 Evaluation Findings

The staff has reviewed the description and evaluation of the TN Eagle-STC containment system and concludes that: (1) the application identifies established codes and standards for the containment system; (2) the package includes a containment system securely closed by a positive fastening device that cannot be opened unintentionally or

45 by a pressure that may arise within the package; (3) the package is made of materials and construction that assure that there will be no significant chemical, galvanic, or other reaction.

The staff has reviewed the evaluation of the containment system under NCT and concludes that the package is designed, constructed, and prepared for shipment so that under the tests specified in 10 CFR 71.71 (NCT) the package satisfies the containment requirements of 10 CFR 71.43(f) and 10 CFR 71.51(a)(1) for NCT with no dependence on filters or a mechanical cooling system.

The staff has reviewed the evaluation of the containment system under HAC and concludes that the package satisfies the containment requirements of 10 CFR 71.51(a)(2) for HAC, with no dependence on filters or a mechanical cooling system.

Based on review of the statements and representations in the application, the NRC staff concludes that the TN Eagle-STC package has been adequately described and evaluated to demonstrate that it satisfies the containment requirements of 10 CFR Part 71, and that the package meets the containment criteria of ANSI N14.5-2014.

5Property "ANSI code" (as page type) with input value "ANSI N14.5-2014.</br></br>5" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..0 SHIELDING EVALUATION The TN Eagle is authorized to transport spent nuclear fuel that is stored in dry storage canisters (DSCs). The TN-Eagle LC is authorized to transport the EOS-37PTH and EOS-89BTH DSCs while the TN-Eagle SC is authorized to transport the 32PT, 32PTH1, 24PT1, 24PT4, FO/FC/FF DSCs including FO, Fuel/Control Components (FC), and FF DSCs.

Several types of BWR fuel assemblies (FAs) with or without fuel channels or PWR FAs with or without control components (CCs) can be contained in the DSCs. The fuel loading conditions are in the following appendices of the application:

DSC Application Appendix for allowable fuel loadings EOS-37PTH 1.6.1 EOS-89BTH 1.6.2 24PT4 1.6.3 32PT 1.6.4 32PTH1 1.6.5 FO, FC, FF 1.6.6 24PT1 1.6.7 5.1 Description of the Shielding Design Packaging Design Features The staff reviewed the general information chapter in the application including the drawings in section 1.5 and determined that the applicant specified all dimensions and tolerances of all components important to the shielding evaluation. The staff also reviewed the additional information on the shielding design in chapter 5 of the application, Shielding Evaluation, for the TN Eagle LC and appendix 5.6.1 of the application, Shielding Evaluation for TN Eagle SC, for

46 the TN Eagle SC. The staff determined that all figures, drawings, and tables describing the shielding features are sufficiently detailed to support an in-depth evaluation.

The staff verified that the drawings contain dimensions with tolerances as well as materials specifications with tolerances for shielding. The shielding design features of the TN Eagle package includes the forged cask body, the lid, shielding rings and the bottom closure plate. The cask is made of forged carbon steel, the primary lid is made of martensitic steel and the bottom closure plate is made of carbon steel. Dimensions of the packaging is provided in table 1-1 of the application. As seen in Figure 1-1 of the application and drawing TN Eagle 01-1100, the forged cask body is encompassed in the steel fitting rings, closed at the bottom with the bottom closure system (which includes the bottom plate), sealed at the top with the primary lid, and then protected with the impact limiters at the top and bottom ends of the cask. The gamma shielding is mostly performed by the forged cask body and shielding ring steel. The gamma shielding at the cask ends is performed by the steel top and bottom assemblies of the cask and axial ends of the DSCs. The neutron shielding is mostly performed by the borated VYAL-B resin blocks in the shielding rings surrounding the body. Neutron shielding is also performed by VYAL-B resin plates at both ends for additional neutron shielding. Shielding is also performed on the fuel basket and inside the impact limiters and adapters. Material properties are provided in section 5.3.2 of the application. Table 5.6.1-15 of the application has important dimensions for shielding.

Summary Table of Maximum Radiation Levels For spent fuel, the applicant analyzed each burnup and cooling time combination required for the loading patterns specified in appendices 1.6.1 through 1.6.7 of the application to ensure that they meet regulatory dose rate limits using the response function method. The applicant calculated the doses using a full cask load with design basis source terms and shows the maximum dose results in table 5-1 of the application for the TN Eagle LC and table 5.6.1-1 of the application for the TN Eagle SC.

The applicant displayed its results for the design basis fuel loadings that produce the highest dose rates at the surface and at 2 m under NCT, and at 1 m under HAC. The applicant additionally calculated the distance from the various surfaces of the TN Eagle LC to meet the dose rate limit for the normally occupied space and showed the results in tables 5-2 of the application.

Table 5-81 of the application shows dose rate contribution for gamma and neutron sources. For TN Eagle SC, table 5.6.1-23 of the application breaks the shows dose rate contribution for gamma and neutron sources for NCT and table 5.6.1-26 of the application breaks down the contributions for HAC.

The staff reviewed tables 5-1 and 5.6.1-1 of the application and found that the calculated dose rates meet the requirements in 10 CFR 71.47 and 10 CFR 71.51. Since the applicant states that the TN Eagle will be operated under exclusive use, the staff verified that the evaluated radiation levels do not exceed those specified in 10 CFR 71.47(b).

The staff verified that the summary table states that the limit of 200 mrem/h will not be exceeded on the external surface of the package. The maximum calculated surface dose rate for the TN Eagle LC is 126 mrem/hr and for the TN Eagle SC is 105 mrem/hr, as seen in table 5.6.1-1 of the application. This meets the regulatory limit in 10 CFR 71.47(b)(1).

47 The applicant showed calculated dose rates at the outer package surface are less than 200 mrem/h, therefore 10 CFR 71.47(b)(2) is met. This regulation requires that the dose rate be limited to 200 mrem/h at the vehicle surface.

The staff verified that the summary table states that the limit of 10 mrem/h will not be exceeded at any point 2 meters from the outer lateral surface of the vehicle. The maximum calculated dose rate at this location was 9.4 mrem/h for the TN Eagle LC and 6.8 mrem/hr for the TN Eagle SC. This meets the requirement in 10 CFR 71.47(b)(3).

In table 5-2 of the application the applicant shows the calculated distance from the package necessary to comply with the 2 mrem/hr limit in the regulation 10 CFR 71.47(b)(4) for normally occupied space for the TN Eagle LC. This meets the regulation by requiring personnel to wear radiation dosimetry at this distance or closer. For TN Eagle LC the applicant shows the minimum distance to meet 2 mrem/hr dose rates and therefore shows that the package meets 10 CFR 71.47(b)(4) at this distance.

The staff verified that the summary table states that the radiation dose rate under HAC does not exceed 1 rem/h at 1 meter from anywhere on the external surface of the package. The maximum dose rate at 1 meter from the surface of the package under HAC is 714 mrem/hr for the TN Eagle LC and 395 mrem/hr for the TN Eagle SC. This meets the requirement of 10 CFR 71.51(a)(2).

5.2 Source Specification The applicant specifies the allowable fuel assembly parameters in appendices 1.6.1 through 1.6.7 of the application. This includes a wide range of allowable spent PWR fuel including control components (CCs), BWR and PWR-MOX assemblies including damaged and failed fuel assemblies. The authorized assemblies for each DSC are summarized below in the following table:

DSC PWR/BWR Max Uranium Loading (kg) per assembly Assemblies Control Components (Note 1)

Damaged /

Failed Fuel Reconstituted Fuel Assembly and / stainless steel rods per reconstituted fuel assembly EOS-37PTH (LC)

PWR 492 Restricted by FQT B&W 15x15, WE 14x14, WE 15x15, WE 17x17, CE 14x14, CE 15x15, CE 16x16, WE 14x14 Yes Maximum 8

damaged, maximum 4 failed fuel only allowed in HLZC1 37 / 5 per Note C of Tables 8.7.1-2 through 30 Requires 2 years additional time for assemblies in the peripheral zones EOS-89BTH(LC)

BWR 198 Restricted by FQT 7x7, 8x8, 9x9, and 10x10 BWR fuel assemblies (with or without channels)

N/A N/A 89 / 5 per Note B of Tables 8.7.2-2 through 10 Requires 6 years additional cooling time in the peripheral zone 24PT4 (SC)

PWR 455.5 CE 16x6 N/A Max 12 24/5

48 32PT(SC) 475kg (per Table 1.6.4-3)

B&W 15x15, WE 17x17, CE 15x15, WE 15x15, CE 14x14, WE 14x14 Yes 28 damaged, 8 failed 32/5 32PTH1(SC)

PWR Per Table 1.6.5-4 490kg (for B&W 15x15 class)

B&W 15x15, WE 17x17, CE 15x15, WE 15x15, CE 14x14, WE 14x14, and CE 16x16 yes 16 damaged 32/5 FO, FC, FF(SC)

PWR 466 FO: B&W 15x15 FC: B&W 15x15 Yes (intact fuel);

must be cooled for 8 years FC:13 FF:

individual fuel cans which confine any gross particles.

No 24PT1(SC)

PWR WE 14x14 Stainless steel clad (up to 4) and MOX 4 damaged.

No The applicant has created fuel qualification tables (FQTs) based on the mass of the assembly, zone within the heat load zone configuration (HLZC) and the loading plan. The loading plans are defined for the TN Eagle LC for the EOS-37PTH and EOS-89BTH in table 5-70 of the application which shows the dose contribution that the FQT is designed to meet for each HLZC zone for each plan.

The TN-Eagle FQTs are summarized below:

49 DSC Condition Tables from the application 492kgU, HLZC1, Plan 1 Zone 1 8.7.1-2 492kgU, HLZC1, Plan 1 Zone 2 8.7.1-3 492kgU, HLZC1, Plan 1 Zone 3 8.7.1-4 492kgU, HLZC1, Plan 2 Zone 1 8.7.1-5 492kgU, HLZC1, Plan 2 Zone 2 8.7.1-6 492kgU, HLZC1, Plan 2 Zone 3 8.7.1-7 492kgU, HLZC2, Plan 1&2 Zone 1 8.7.1-8 492kgU, HLZC2, Plan 1&2 Zone 2 8.7.1-9 492kgU, HLZC2, Plan 1 Zone 3 8.7.1-10 492kgU, HLZC2, Plan 1 Zone 4 8.7.1-11 492kgU, HLZC2, Plan 1 Zone 5 8.7.1-12 492kgU, HLZC2, Plan 1 Zone 6 8.7.1-13 492kgU, HLZC2, Plan 2 Zone 3 8.7.1-14 492kgU, HLZC2, Plan 2 Zone 4 8.7.1-15 492kgU, HLZC2, Plan 2 Zone 5 8.7.1-16 492kgU, HLZC2, Plan 2 Zone 6 8.7.1-17 400kgU, HLZC1, Plan 1 Zone 1 8.7.1-18 400kgU, HLZC1, Plan 1 Zone 2 8.7.1-19 400kgU, HLZC1, Plan 1 Zone 3 8.7.1-20 450kgU, HLZC1, Plan 1 Zone 1 8.7.1-21 450kgU, HLZC1, Plan 1 Zone 2 8.7.1-22 450kgU, HLZC1, Plan 1 Zone 3 8.7.1-23 (Exceptions) 492kgU, HLZC1, Plan 1 Zone 1 8.7.1-24 (Exceptions) 492kgU, HLZC1, Plan 1 Zone 2 8.7.1-25 (Exceptions) 492kgU, HLZC1, Plan 1 Zone 3 8.7.1-26 (Exception-counterparts) 492kgU, HLZC1, Plan 1 Zone 1 8.7.1-27 (Exceptions-counterparts) 492kgU, HLZC1, Plan 1 Zone 2 8.7.1-28 EOS-37PTH (Exceptions-counterparts) 492kgU, HLZC1, Plan 1 Zone 3 8.7.1-29 198kgU, HLZC1, Plan 1, Zone 1 8.7.2-2 198kgU, HLZC1, Plan 1, Zone 2 8.7.2-3 198kgU, HLZC1, Plan 1, Zone 3 8.7.2-4 (Exceptions) 198kgU, HLZC1, Plan 1, Zone 1 8.7.2-5 (Exceptions) 198kgU, HLZC1, Plan 1, Zone 2 8.7.2-6 (Exceptions) 198kgU, HLZC1, Plan 1, Zone 3 8.7.2-7 (Exceptions-counterparts) 198kgU, HLZC1, Plan 1, Zone 1 8.7.2-8 (Exceptions-counterparts) 198kgU, HLZC1, Plan 1, Zone 2 8.7.2-9 EOS-89BTH (Exceptions-counterparts) 198kgU, HLZC1, Plan 1, Zone 3 8.7.2-10 32PTH1, 32PT, 24PT4, 24PT1, FC/FO/FF FQT for DSCs inside TN Eagle SC 8.7.3-1 For the TN Eagle LC, EOS-37PTH DSC control components are allowed. The applicant shows the allowable Co-60 source in table 1.6.1-17 of the application. For control components that have source nuclides other than Co-60 the applicant states that users can determine if the source provides the dose equivalent to Co-60 using the response functions as discussed in section 5.4.1.5 of the application. For fuel assemblies that contain control components, that are loaded in a peripheral zone of HLZC1 or HLZC2, users need to add additional cooling time to the fuel assembly as required by table 8.7.1-30 of the application.

50 For the EOS-37PTH and the EOS-89BTH baskets, the TN Eagle is allowed to transport fuel assemblies with a shorter cooling time than those required in the standard FQTs. This is allowed for only these two baskets, and only fuel assemblies being loaded according to the FQTs for 492 kg U (EOS-37PTH) and 198 kg U (EOS-89BTH). An exception is allowed if it has a cooling time that meets one of the exception tables in the application (8.7.1-24 through 26 for PWR and 8.7.2-5 through 8.7.2-7 for BWR).

If an exception is loaded, it must be adjacent to at least two fuel assemblies that have the corresponding lower contribution to dose rates defined in the counterpart tables in table 8.7.1-27 through 8.7.1-29 for PWR or 8.7.2-8 through 8.7.2-10 for BWR. According to the information provided by the applicant, the staff found this conservative and acceptable as the additional source term from an exception would be balanced.

For the TN Eagle SC, control components are allowed in the 32PT, 32PTH1 and the FC/FO/FF DSCs. The allowable CC source terms for the 32PT and 32PTH1 are located in tables 1.6.4-4 and 1.6.5-3 of the application, respectively. The staff verified that the total allowable source is consistent with the source term assumed for these components in table 5.6.1-9 of the application. The source term used in table 5.6.1-9 has an axial dependency in that the source has been distributed along the entire length of the fuel. This is reflected in the allowable source term for CCs in tables 1.6.4-4 and 1.6.5-3 of the application in that TPAs and ORAs, which would have significant source near the top are restricted to the source term in table 5.6.1-9 of the application for the plenum and top.

Spent Fuel Source Term As discussed in section 5.2.1 of the application, the applicant calculated the gamma and neutron source term from radioactive fission products using the TRITON and ORIGEN-ARP modules of the SCALE 6.0 system using the ENDF/B-VII 238-group cross section library.

ORIGEN-ARP uses interpolated cross sections that can be generated using the TRITON module of the SCALE code system to simulate reactor conditions. ORIGEN is considered acceptable to the staff per the guidance in section 3 of NUREG/CR-6802, Recommendations for Shielding Evaluations for Transport and Storage Packages. TRITON is a transport theory-based code, mentioned in section 3.2.6 of NUREG/CR-6802 as acceptable for this type of analysis. It represents more advanced and detailed 2-D reactor physics solution method as compared to SAS2H (which is recommended and used in NUREG/CR-6802) and the staff found the use of this code acceptable for the TN Eagle.

Spent Fuel Assemblies The applicant selected a PWR 15x15 fuel assembly as the design basis assembly for generating PWR spent fuel assembly source terms for the EOS-37PTH DSC. The staff found this appropriate as the mass is the only assembly characteristic that made the greatest difference in dose as documented in NUREG/CR-6716, Recommendations on Fuel Parameters for Standard Technical Specifications for Spent Fuel Storage Casks (NRC, 2001). The staff verified that the uranium mass assumed for the assembly is consistent with the maximum allowable uranium for this DSC from table 1.6.1-1 of the application and found the selection of this assembly acceptable.

The applicant selected the BWR 7x7 fuel assembly as the design basis assembly for generating BWR spent fuel source terms for the EOS-89BTH. The staff found this appropriate based on the above discussion and verified that the uranium mass assumed for the assembly is consistent

51 with the maximum allowable uranium for this DSC from table 1.6.2-1 of the application and found the selection of this assembly acceptable.

For the TN Eagle SC the applicant modeled the 32PTH1 basket to represent all of the other allowable baskets. The staff compared all of the contents for other allowable baskets (32PTH1, 32PT, 24PT4, 24PT1, FC/FO/FF) within the TN Eagle SC and found that the 32PTH1 is bounding for the following reasons:

The 32PTH1 contents bounds all allowable assemblies within the other baskets.

The 32PTH1 has at least as many assemblies as any other basket allowed within the TN Eagle SC.

Based on the loading table, table 8.7.3-1 of the application, the 32PTH1 has the highest allowable burnup and lowest minimum cooling time.

The shielding properties of the 32PTH1 were determined to be bounding by the staff as discussed in this SER.

The assembly chosen for the 32PTH1 is the same as the EOS-37PTH. The staff found this acceptable because similar to the EOS-37PTH, it has the largest mass of all of the allowable assemblies for all of the TN Eagle SC baskets (see explanation above). There are other contents allowed in the 24PT1 that have unique composition that cannot be directly compared to typical PWR fuel and the assessment in NUREG/CR-6716 may not be applicable. This is UO2 stainless steel-clad fuel and MOX fuel.

Stainless steel-clad fuel can have a significant gamma source due to the activation of Co-59 impurity into Co-60 within the cladding and the MOX fuel would have different spent fuel nuclides that may not be bounded by UO2 fuel. The applicant did not perform evaluations for these two fuel types allowed for transport within the 24PT1 and the staff found this acceptable for the following reasons:

The 24PT1 basket has fewer assemblies than the design bass 32PTH1 and should have a reduced source term.

The maximum allowable burnup for the 24PT1 (45 GWd/MTU for UO2 and 25 GWd/MTU for MOX) is lower than the 32PTH1 (62 GWd/MTU). Although the spent fuel nuclides will not be the same as 32PTH1 fuel due to the presence of MOX fuel, with more burnup, and lower cooling time (22.7 years versus 39 years) and higher enrichment for UO2 (3.12% versus natural uranium for UO2, the Pu enrichment vector for MOX cannot be compared directly), it is the staffs judgment that the PWR fuel analyzed in the 32PTH1 would bound the source term for the allowable MOX fuel.

The minimum cooling time for the 24PT1 is 39 years for all assemblies, this should be long enough for any significant amount of Co-60 within the stainless-steel cladding to have decayed. With a half-life of about 5.3 years, the minimum decay time is more than 7 half-lives for Co-60.

Depletion Characteristics Reactor Operating Parameters for Depletion Calculations To perform the fuel depletion analysis, the applicant states some of the reactor operating parameters it used in section 5.2.2 of the application. As shown in NUREG/CR-1617, a higher

52 specific activity is conservative for performing dose rate evaluations. Based on appendix B of NUREG/CR-6802, the staff found that the specific power chosen by the applicant is relatively high and is acceptable to the staff. The applicant chose a water density of 0.63 g/cm3 and 0.432 g/cm3 for PWR and BWR fuel respectively. The applicant stated that these are lower than typical reactor operating values and result in larger neutron source terms due to the harder neutron spectrum during irradiation.

BWR fuel experiences a large variation in water density (void fraction) along the axial height of the core as it enters the core as liquid water and exits the core as nearly all vapors. The applicant did not calculate any uncertainty related to assuming a uniform void fraction over an axially varying profile.

However, the staff performed independent calculations using Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) which does account for axial variations in void profile and found that the BWR configurations that it modeled meet regulatory dose rate limits.

The staffs model did show larger differences at the top and bottom of the package and this could be due to the assumptions related to the axial void profile. As these surfaces had more margin to the regulatory limit than the location 2 meters radially, the staff did not investigate whether the difference was due to the assumptions related to the axial void profile. Therefore, staff has reasonable assurance that the assumption of a uniform axial void profile is acceptable for BWR fuel for the TN Eagle.

In section 5.4.1.2 of the application, the applicant states that it created ORIGEN-ARP models for each combination of burnup, enrichment and cooling time in the FQTs shown in chapter 8. The applicant used these to develop the appropriate FQTs using the response functions. The use of the response functions has limitations in that the applicant did not have a comprehensive dose rate tally associated with them.

To benchmark the response function method, the applicant used a few design basis sources to calculate dose rates using a full TN Eagle cask model. The applicant states that PWR design basis sources are presented in table 5-9 through 5-18 of the application. The BWR design basis sources are presented in table 5-19 through 5-21 of the application.

The staff was unable to determine all the depletion parameters used by the applicant within the analysis such as boron concentration (PWR), presence of burnable absorbers or control blades, fuel temperature, etc. As discussed in section 5.5.1 of this SER, the staffs calculations of source term, comparing directly to the applicant is higher. The staff did not investigate the discrepancy as its overall calculations (using source terms it determined are generated using conservative depletion parameters) did show that the system meets dose rate limits.

Therefore, the staff determined that either the differences in source term are not significant, or there are other conservative assumptions within the applicants analysis that are compensating for this uncertainty and that overall, the system meets regulatory limits.

Gamma Source from SNF The applicant calculates the gamma source from ORIGEN-ARP as discussed in the previous section of this SER and represents it using the 18-group structure shown in the design basis source tables 5-9 through 5-18 of the application and limits the response function method to 7 groups that span the range discussed below. The staff found that it spans the energy range

53 discussed in section 5.4.2.2 of NUREG-2216 stated to contribute to external dose rates. The group structure does not conform exactly to the recommendations in section 3.3.3.3 of NUREG/CR-6802 with respect to the location of the energy bins, however, it is the staffs judgment that this should not have a significant effect on calculated dose rates as some gamma energies may be represented by a lower energy while some may be represented by a higher energy and its the staffs judgment that the group structure is capable of representing an average source term. The staff also used its confirmatory calculations using UNF-ST&DARDS (discussed in this SER) which uses a very detailed group structure to determine that the group structure developed by the applicant does not have a significant impact on dose rates.

In section 5.6.1.2.7 of the application the applicant states that it runs the MCNP code in the coupled neutron-photon mode to account for secondary gammas. The staff verified that the appropriate cards were used in the sample MCNP input files and found that secondary gammas would be generated and are appropriately accounted for using MCNP. The staff found this acceptable.

As discussed in section 5.4.1.1 of the application, within the response function method employed by the applicant, it included all most gammas in the spectrum. The staff found this acceptable for this application as independent staff calculations show that the gamma source strength from spent fuel above the upper limit of MeV is at least one order of magnitude smaller than the gammas at or below the limit of MeV, and gammas below the bottom limit of MeV would not penetrate the cask shielding. This conclusion is consistent with the principle of radiation absorption as reflected in the flux-to-dose rate conversion factors for gammas, i.e.,

high energy gammas will not be absorbed by human organs as effectively as lower energy gammas and the gammas with energy below the bottom limit of MeV will not penetrate the layers of shielding materials of the package.

In addition, since the applicant also calculated the dose rate using the full energy spectrum and a full model of the cask to confirm the dose rates are below regulatory limits using a small set of design basis sources to confirm the dose rates. The staff also performed independent calculations using UNF-ST&DARDS that use a gamma energy group structure from 0.01 MeV up to 10 MeV to confirm package dose rates (see discussion in this SER).

Neutron Source from SNF The applicant uses the spectra from the Cm-244-watt fission spectrum to represent the neutron source from spent fuel. The staff found this acceptable based on the analyses in NUREG/CR-6700, Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms Related to Transport and Interim Storage of High-Burnup LWR Fuel, section 5, the neutron source is primarily from spontaneous fission and Cm-244 is the primary contributor to dose rate coming from neutrons.

The applicant used the ORIGEN-ARP code to calculate the magnitude of the neutron source.

The staff performed independent calculations as discussed in this SER, of the source term and was able to verify the magnitude of the source is consistent with what was calculated by the applicant.

To account for subcritical multiplication, as discussed in section 5.2.3 of the application, the applicant increased the neutron source to account for subcritical multiplication. The staff found this is consistent with the guidance in section 4.2 of NUREG/CR-6802. The applicant discusses how it calculated the keff in section 5.2.3 of the application and presents the keff values in tables 5-25 and 5-26 of the application for PWR and BWR fuel, respectively. The staff did not verify

54 these values however it performed independent calculations of the dose rate using UNF-ST&DARDS (discussed in section 5.5 of this SER) which uses a different method for accounting for secondary neutrons from subcritical multiplication.

The method used in the UNF-ST&DARDS calculations uses the energy spectra directly from the ORIGAMI calculations. It also models a pin-by-pin geometry of the UO2 rods for more accurate subcritical multiplication. The neutron reactions that account for sub-critical multiplication is part of the neutron physics tracked within the MAVRIC calculations employed by the UNF-ST&DARDS code.

Although the staff did not compare whether the methods for neutron propagation used by the applicant and the staff are directly comparable, the staff has reasonable assurance that the contents, when modeled as a whole (which includes gamma contributions) meet regulatory dose rate limits, given the different methods of evaluating neutron dose, the staff found this gives assurance that the applicants method for calculating sub-critical multiplication is reasonable.

Axial Burnup Profile PWR Burnup Profile The applicant states that it used the axial burnup profile from NUREG/CR-6801, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses, for the 26-30 GWd/MTU burnup range. Lower burnup assemblies have a higher peaking which is conservative for use in the shielding evaluations as it maximizes dose at the peak. Although this NUREG was created to investigate the effect of burnup profiles on criticality (keff) evaluations, the staff found that this should be relatively conservative for a shielding evaluation as well as burnup profiles that were found to be more reactive (higher keff) have a larger end effect, meaning less burnup at the ends, which would mean there is higher peaking in the middle of the assembly. Higher peaking is more conservative for shielding evaluations, therefore the staff found that it could also conclude that burnup profiles that are conservative for keff evaluations are also conservative for dose rate calculations and therefore the staff found that the burnup profile selected for PWR fuel is reasonable and acceptable.

The applicant varied the gamma source proportionally with the axial burnup profile and varied the neutron source by a power of 4.0 to 4.2. The staff found this appropriate as it is consistent with section 3.3.1 of NUREG/CR-6802.

In addition, the staff performed confirmatory calculations, as discussed in section 5.5 of this SER, using the UNF-ST&DARDS code and it uses conservative burnup dependent axial burnup profiles for the PWR spent fuel (Radulescu, Banerjee, LeFebvre, Miller & Scaglione, Shielding Analysis Capability of UNF-ST&DARDS, Nuclear Technology, September 1, 2017, https://doi.org/10.1080/00295450.2017.1307643). The staff was able to confirm that the package meets the regulatory dose rates using this code and therefore found that this provides additional assurance that the applicants assumed burnup profile is appropriate.

BWR Burnup Profile The applicant stated in section 5.2.3 of the application that it used the BWR burnup profile from the reference: Design Data Document DI-81001-02, NOK Document, Technical Specification for the Supply of Transportable Casks for the Storage of Kernkraftwerk Leibstardt (KKL) Spent Fuel in ZWILAG, TS 07/01, Rev. 1.

55 The staff is not familiar with this document and did not review it to determine whether the burnup profile used is appropriate. However, burnup profile has a more significant impact at the surface of the cask, as a very peaked profile may create a localized area of higher dose. The TN Eagle is limited by the 2-meter dose rate and at that distance, burnup profile has less of an impact on dose rates, therefore the staff determined that the burnup profile used is of relatively low significance. In addition, the staff was able to perform independent calculations using UNF-ST&DARDS (discussed in section 5.5 of this SER) which uses conservative burnup profiles based on burnup (Radulescu, Banerjee, LeFebvre, Miller & Scaglione, Shielding Analysis Capability of UNF-ST&DARDS, Nuclear Technology, September 1, 2017, https://doi.org/10.1080/00295450.2017.1307643).

Spent Fuel Activated Structural Materials The applicant states in section 5.2.2 of the application that it used the composition of structural materials from Oak Ridge National Laboratory (ORNL)/TM-11018, Standard and Extended Burnup PWR and BWR Reactor Models for the ORIGEN2 Computer Code. The primary sources of activity in the structural materials of the fuel assemblies are from Co-60 which is created mainly by neutron activation of Co-59 in the steel and Inconel material. The composition of the other fuel structural materials generally does not contribute much to external dose so the staff found the composition from ORNL/TM-11018 acceptable for this purpose. The Co-59 impurity is the most important impurity to characterize as it gets activated into Co-60 and can provide a significant contribution to dose. The applicant states that except 24PT1 DSC, for old fuel with higher cobalt impurity in materials (Stainless Steel 304, Inconel-718, and Inconel X-750) than shown in table 5-5 of the application, add 13.0 years of additional cooling time.

The mass of the various elements making up the non-fuel regions of the assembly used for Co-60 activation calculations are in tables 5-7 and 5-8 of the application. The staff used U.S.

Department of Energy (DOE)/RW-0184, Volume 3 of 6, Appendix 2A, Physical Descriptions of LWR Fuel Assemblies, December 1987, to verify the hardware masses for the different zones for the spent fuel hardware and found the assumed values in tables 5-7 and 5-8 of the application are consistent and therefore acceptable to the staff.

The applicant calculated the activity of the Co-60 from the Co-59 activation using the ORIGEN-ARP using the in-core region flux at full power. The applicant modified the mass for each region by using scaling factors in table 5-6 of the application. These are from PNL-6906, Spent Fuel Assembly Hardware: Characterization and 10 CFR 61 Classification for Waste Disposal, June 1989. The process of using the scaling factors is discussed in NUREG/CR-6802, Recommendations for Shielding Evaluations for Transport and Storage Packages, May 2003, (ADAMS Accession No. ML031330514). PNL-6906 is referenced in section 5.5.2.1 of NUREG-1617 and also in section 3.3.2 of NUREG/CR-6802.

On these bases, the staff found that the applicants approach meets the acceptance criteria as specified in NUREG-2216 and is acceptable for determining source term from activated fuel hardware.

Control Components The TN Eagle is authorized to transport control components (CCs) which are defined by the applicant as any component that has been placed within a PWR reactor fuel assembly instrument and guide tubes during or after operation and includes items such as: burnable poison rod assemblies (BPRAs), TPAs, CRAs, control element assemblies (CEAs), Control

56 Spiders, RCCAs, APSRAs, ORAs, PPSAs, VSIs, neutron source assemblies (NSAs), Neutron Sources, ITTRs or Anchors, Guide Tube Inserts, and BPRA Spacer Plates.

The allowable CCs for each DSC is summarized in section 5.2 of this SER.

For the EOS-37PTH DSC CCs are allowed in every location and are restricted to a maximum amount of Co-60 per zone. The fuel assemblies in peripheral locations loaded with CCs are subject to additional cooling time restrictions. NSAs are restricted to interior locations of the basket.

Section 5.2.4 of the application discusses the groups of control components such as BPRA and TPA. The staff did not verify the assumptions related to the mass or activation of these components as they are limited by the Co-60 content which was used in the dose rate evaluations to demonstrate that the package meets regulatory limits. As long as the component that is loaded into the cask meets the Co-60 limits then the mass and activation assumptions used to create that amount of Co-60 is irrelevant.

The applicant used the source term from table 5-32 of the application to represent the CCs within the TN Eagle. The staff performed a study comparing this source to the allowable Co-60.

The staff used Microshield 12 using iron roughly equivalent to the amount of steel shielding through the TN Eagle package to compare dose rates for both source terms and found that the more detailed source term in table 5-32 gives higher dose rates and is therefore conservative to represent the source of these components which is only limited by Co-60.

Although this is appropriate for activated metals resulting from neutron activation, for which Co-60 is the dominant nuclide contributing to dose, it may not be appropriate to represent sources from control rods made of Hafnium or Ag-In-Cd. In addition, it may not appropriate for NSAs or other neutron sources. To account for the source from these components, the applicant states that up to 2 NSAs can be loaded in the interior compartments represented by locations in Zone 1 of HLZC1 for the EOS-37PTH DSC, and the number of NSAs is limited compared to 37 FAs loaded in the EOS-37PTH DSC, therefore the impact of the NSAs on the shielding performance of the cask may be neglected, and dedicated shielding models with the neutron sources from the NSAs are not needed.

Irradiated Stainless Steel Rods As discussed in this SER, the TN Eagle is authorized to ship reconstituted fuel assemblies and that includes assemblies with irradiated replacement stainless steel rods. These are steel rods that would replace a fuel rod that has failed in an assembly and the assembly is burned again within the reactor core, therefore irradiating the steel rod. These steel rods become activated producing a gamma source term primarily due to the production of Co-60 from Co-59 impurities.

The contribution to dose rate may be balanced somewhat due to the replacement of an actual fuel rod, however when determining the overall impact on dose rate the applicant should also consider the change in self-shielding when spent UO2 is replaced with a steel rod. Increased Co-60 source can have a significant impact on dose especially for fuel assemblies with a lower cooling time.

All assemblies allowed within the TN Eagle (SC or LC) to have irradiated stainless-steel rods have a limit of five per assembly.

There are no restrictions on where reconstituted assemblies are placed within the DSC. The applicant discusses the assumptions it used to determine the source from the irradiated

57 stainless-steel rods in section 5.2.5 of the application for the LC and section 5.6.1.2.5 of the application for the SC, although the assumptions are the same for both. The applicant used ORIGEN-ARP and reduced the allowable mass equivalent to five stainless steel rods, and reduced the overall power, but kept the specific power the same thereby extending the burnup.

The staff verified the applicants mass of stainless-steel rods using assembly design information from table 6.8.3-2 of the SAR and DOE/RW-0184 Volume 3 of 6. The staff found that these are reasonable assumptions to determine the source term for an assembly with irradiated steel replacement rods present.

The applicant determined that the dose rate contribution from reconstituted rods in the inner zones is negligible. Based on staff calculations (see this SER) the staff also found that the inner zones have less contribution to the overall regulatory compliance of the package, given it is limited by the 2-meter radial dose and there are large margins to the regulatory limits at the top and bottom of the package. However, the applicant states that for all reconstituted fuel loaded in the peripheral zones shown in Figure 5-4 and Figure 5-5 of the application for the EOS-37PTH DSC and the EOS-89BTH DSC, respectively, additional cooling time are required for the EOS-37PTH DSC and the EOS-89BTH DSC, respectively, according to table 5-75 of the application.

No additional cooling time is required for loading reconstituted fuel into the TN Eagle-SC. Based on the large margins to the regulatory limit for this variant, the staff found this acceptable.

The staff verified that Note C and Note B of the FQTs located in appendices 8.7.1 and 8.7.2 of the application for the EOS-37PTH and EOS-89BTH, respectively, include the requirement for additional cooling time for reconstituted fuel. The staff found that this adequately accounts for the additional source term from reconstituted fuel and found it acceptable.

Staff does not know the impurity level of cobalt within the steel components assumed by the applicant, however it performed (see this SER) independent calculations assuming additional cobalt in the amount of 800ppm of 2/3 of steel rod mass and found that for the calculations it ran that the additional cooling time of 2 years for the EOS-37PTH and 6 years for the EOS-89BTH was enough to compensate for any increase in dose due to the presence of irradiated stainless steel rods within reconstituted fuel.

For the canisters allowed in the TN Eagle SC that allow irradiated replacement rods, they do not require any additional cooling time. The staff also performed an independent calculation increasing the cobalt within the 32PTH1 in the TN Eagle SC Type C and found that due to the large margins to the regulatory limit, considering the increase in dose rate due to these components the TN Eagle SC is still within regulatory dose rate limits.

5.3 Model Specification The applicant modeled the EOS-37PTH and EOS-89BTH using the dimensions in table 5-35 of the application. The staff compared these dimensions with those from drawings and found that they are consistent with the dimensions in Drawings EOS01-71-1001.

The TN Eagle LC and SC cask bodies are modeled with minimum dimensions or lower dimensions than drawing thicknesses. This is conservative and acceptable. The shielding rings are modeled with lower density than carbon steel, this is conservative and acceptable.

The TN Eagle SC has two variants of the shielding rings. The total thickness of the rings for the two variants is the same. The applicant chose to do all analyses with the Type B ring geometry.

The staff did not find this selection to be obviously bounding as different allowable contents may have more gamma or more neutron radiation contributing to dose depending on its

58 characteristics (burnup/enrichment/cooling time), and therefore the Type B or the Type C may be more limiting based on the loading. However, the applicants calculations (table 5.6.1-1 of the application) show that there are large margins to the regulatory limit for the SC.

Staff calculations confirm that there are large margins to the regulatory limit. The staff also performed calculations using both the Type B and Type C shielding ring configurations. Staff calculations show that although there are slight differences in dose rates for the Type B and C depending on the loading, the difference in dose rate between the two is small enough that it does not affect the ability for these configurations to meet regulatory dose rate limits and therefore the staff found it acceptable for this application for the applicant to use the Type B shielding rings for the TN Eagle SC configurations.

For the TN Eagle LC, the applicant modeled the EOS-37PTH and EOS-89BTH baskets explicitly. The staff found this conservative and therefore acceptable. In addition, the applicant modeled minimum thicknesses for the EOS-37PTH and EOS-89BTH and in some cases neglected for the EOS-89BTH. This is conservative and therefore acceptable.

For the TN Eagle SC, the applicant modeled the 32PTH1 to bound all other allowable canisters.

Based on the comparison of allowable contents for all of the canisters allowed in the SC (32PTH1, 32PT, 24PT4, 24PT1, FC/FO/FF), considering factors such as maximum uranium mass per assembly, number of assemblies, maximum burnup, minimum cooling time, the staff determined that this basket would be the most bounding for all of those allowed within the TN Eagle SC, and found the 32PTH1 fuel to be the bounding source (see section 5.2.1.1 of this SER).

In addition, the shielding characteristics of the 32PTH1 are also bounding with respect to the other allowable canisters within the TN Eagle SC. The 32TPH1 has slightly less material at the cask side. Based on table 5.6.1-16 of the application, the top and bottom of the canisters cannot be easily compared as the 32PTH1 has less steel than other canisters; however, some have lead shielding, and it is not easy to infer which would have better shielding capabilities. Still the baskets with lead, are thick enough that it is within the staffs judgment that the 32PTH1 is the most bounding (i.e., represents the canister with the least amount of shielding). Also, the limiting dose rate for the TN Eagle is the 2-meter dose rates at the side of the cask and the staff does not consider the top or bottom to be as significant in terms of determining regulatory compliance.

Normal Conditions of Transport The applicants model that evaluates the dose rates around the TN Eagle package under NCT consists of hypothetical reconfiguration of HBU fuel, and effects on the neutron shield resin due to high temperature.

Hypothetical Accident Conditions As discussed in the application, the applicants model used to evaluate the dose rates around the TN Eagle package under HAC takes into account 100 percent fuel failure and partial resin loss.

Configuration of Source and Shielding As discussed in this SER, the applicant uses an MCNP to model the TN Eagle package to calculate external dose rates. Since this is a 3-dimensional code capable of modeling complex

59 geometries, the TN Eagle package can be modeled explicitly without gross approximations.

The staff compared the modeled dimensions of the components as listed in tables 5-35, 5-36 of the application and found that these dimensions were consistent with the package drawings presented in section 1.3 of the SAR. The applicant used the minimum dimensions for all the package components expected to have a significant effect on dose rates.

Spent Nuclear Fuel and Hardware Modeling The applicant homogenized the fuel assemblies rather than model them explicitly (i.e., pin by pin) and placed the homogenized fuel assemblies within the basket which was modeled explicitly. The staff found that modeling fuel assemblies as homogenous rather than exact geometry is acceptable based on discussion in section 4.2 of NUREG/CR-6802 which states that explicit pin by pin modeling produces statistically equivalent results as homogenized fuel assemblies.

The applicant also modeled the bottom nozzle, plenum, and top nozzle portions of the assembly. The applicant shows the axial lengths of the fuel assembly components in tables 5-3 and 5-4 of the application for PWR and BWR fuel, respectively. The staff verified these dimensions using information on the B&W 15x15 from fuel design information from DOE/RW-0184, Volume 3 of 6, appendix 2A, Physical Descriptions of LWR Fuel Assemblies, December 1987. There are slight differences between the applicants dimensions and those from DOE/RW-0184; however, the staff performed an evaluation using Microshield Version 12.11X and a Co-60 source to determine that the dimensions assumed by the applicant, considering the homogenized density are represented conservatively with respect to shielding capability. The applicant shows the homogenized densities in tables 5-40 and 5-41 of the application for PWR and BWR fuel, respectively. The staffs evaluation of the materials used in the applicants model is discussed in this SER.

For PWR fuel, the applicant has ignored any of the materials associated with the control components. For BWR fuel, the applicant has ignored the mass of the channel. This is a small amount of material as compared to the fuel assemblies and although these materials may be a source for secondary gammas, secondary gammas are more likely to be generated with lower energy neutrons. Neglecting this additional material close to the source is likely conservative as additional materials may contribute to additional shielding therefore the staff found neglecting control component and channel materials acceptable.

Under HAC, the applicant modeled the failure of all fuel using the failed fuel models discussed in section 5.3.1.1 of the application and in the following section of this SER.

Damaged and Failed Fuel Modeling The EOS-37PTH is allowed up to eight damaged fuel assemblies or four failed fuel assemblies when in HLZC1. Based on its definition within the application, damaged fuel will maintain its configuration under NCT, so it is modeled the same as intact fuel. The staff found this acceptable.

There are four failed fuel assemblies allowed within the EOS-37PTH. The applicant does not model the FFC. Although there could be secondary gammas generated within the additional steel from the FFC, based on the (n,) cross section of iron, secondary gammas are more probable from neutrons that are at lower energies so additional steel on the interior of the package would more likely reduce dose rates due to providing extra gamma shielding rather

60 than produce secondary gammas as the neutrons near the fuel assembly are higher energy.

The staff found neglecting the FFCs to be conservative.

In section 5.1 of the application, the applicant states that design basis gamma sources are used for failed fuel with altered axial source distribution as described in section 5.3.1.1 of the application.

Fuel/Source Modeling for HBU Fuel The applicant discusses its approach for addressing considerations related to HBU in section 5.3.1.1 of the application. HBU is defined as fuel experiencing burnup above 45 GWd/MTU. To address the uncertainty in the mechanical performance of the fuel assembly for high burnup fuel, the applicant performed calculations simulating reconfiguration consistent with the recommendations in NUREG-2224, Dry Storage and Transportation of High Burnup Spent Nuclear Fuel. The staff found that these scenarios are appropriate to address reconfigured fuel based on the analyses in NUREG/CR-7203.

To account for any increase in dose rates due to reconfigured fuel, the applicant added 3 years cooling time to the source terms for this fuel as shown in tables 5-96 through 5-101 of the application for the TN Eagle LC. The staff verified that all the FQTs in sections 8.7.1, 8.7.2, and 8.7.3 of the application have the requirement within the notes section for the tables that HBU fuel that has been in dry storage for more than 20 years must have an additional 5 years cooling time. The staff found that this is conservative because used to demonstrate that the package meets regulatory dose rate limits with reconfigured fuel.

As discussed in section 5.4.4 of the application, the results of the applicants reconfiguration studies show that there are some localized increases in dose rate at the surface due to the relocation of fuel, but there is an overall reduction of dose rate at 2 meters. The staff found that the applicant has addressed the possibility of reconfiguration from HBU fuel conservatively and consistent with NUREG/CR-2224 and found the applicants reconfiguration analyses acceptable.

Material Properties The applicant discusses the material properties it used within the dose rate evaluations in section 5.3.2 of the application. The applicant stated that it obtained material properties for basic materials such as steel and aluminum from Pacific Northwest National Laboratory (PNNL)-

15870. These are listed in table 5-37 of the application. The staff found that this is acceptable for standard materials.

Packaging Materials The EOS-37PTH and the EOS-89BTH can have basket walls made form metal matrix composite (MMC) or Boral for criticality control. The applicant chose to model this as MMC for the shielding evaluation. Considering these materials are relatively close in composition and dont provide significant shielding, the staff found the choice of MMC as acceptable. For the EOS-37PTH and EOS-89BTH the applicant modeled the MMC, which is a non-standard material, the applicant used the material properties in table 5-38 of the application. The staff compared these compositions to the allowable material specification for these materials and found that they are modeled conservatively or consistent with these specifications.

61 The composition of the VYAL-B used in the neutron shielding rings within the dose rate calculations is shown in section 9.1.7 of the application. The staff verified that this was consistent with the material data sheet (Ref. DI/RI-A-1-5-02 Rev. 01) for this material. The staff found this conservative with respect to the results in R&DDI001-02-B-17 rév.01, Note on the Fire Resistance of the VYAL-B Resin. This is conservative and acceptable to the staff.

SNF Materials The applicant modeled the fuel and fuel hardware components as a homogenous mixture using the composition from tables 5-40 and 5-41 of the application for PWR and BWR fuel, respectively. The composition of the fuel hardware is located in tables 5-3 and 5-4 of the application. The staff verified the composition of the fuel hardware as well as the homogenized density using the fuel assembly design information from DOE/RW-0184 Volume 3 appendix 2A.

For the active fuel, the applicant assumed 3% enriched UO2. The staff found this acceptable as the applicant suppressed fission using the NONU card in MCNP and performs separate k-eff calculations to account for sub-critical multiplication therefore the enrichment is not as important as it would be if the sub-critical multiplication were calculated by the transport calculation. The applicant modeled the spent fuel composition as fresh UO2. The staff found that this is acceptable, as studies performed (G. Radulescu (ORNL), P. Stefanovic (ORNL), ORNL/SPR-2021/2373, A Study on the Characteristics of the SNF Radiation Source Terms of Spent Fuel and Various Non-Fuel Hardware Sensitivity Studies for Shielding Applications, dated January 5, 2022) show that there is no significant difference in dose rate between using fresh fuel nuclides and spent fuel nuclides.

The results of these studies show that using fresh fuel nuclides may be conservative due to the additional neutron source from subcritical multiplication. Given the results being so similar, the staff found that assuming fresh fuel is acceptable for this model. In addition, the staff used the UNF-ST&DARDS code to perform independent calculations and was able to determine that the package met regulatory limits using a sampling of allowable loadings. The UNF-ST&DARDS code models the spent fuel nuclides explicitly and therefore the staff determined modeling the spent fuel as fresh UO2 acceptable.

Failed and Reconfigured Fuel Materials The materials for the failed and reconfigured fuel are nearly the same as for intact fuel.

Compositions have changed slightly; however, the staff does not believe it is enough to make a significant difference in dose rates, especially given the conservative nature of these analyses.

The density has changed to represent the different geometries and relocation of the fuel. The geometry and density for the failed fuel are represented in tables 5-43 and 5-44 of the application for PWR and BWR fuel respectively. This is based on the postulated representations of failed fuel the staff found the dimensions and densities of the failed fuel to be appropriate.

5.4 Evaluation Codes The applicant calculated the gamma and neutron source terms from radioactive fission products using the ORIGEN-ARP module of SCALE 6.0 system using the ENDF/B-VII 252-group library provided by the developer of the code (ORNL). The applicant stated that it used TRITON to develop models of the design basis fuels and referenced the NUHOMS EOS System Updated FSAR, Docket Number 72-1042, Revision 3 which states that it used TRITON to develop cross section data using the fuel parameters for the B&W 15x15 design data in chapter 2 of that

62 report. It states that the default cross sections were used for the GE 7x7 representing BWR fuel.

The staff found the use of the TRITON code to generate cross sections and ORIGEN-ARP for calculating depletion decay as acceptable codes. ORIGEN-ARP is discussed in NUREG/CR-6802 and has been found acceptable by the staff for this purpose. TRITON is a newer code and has much more detailed physics and is acceptable to the staff for modeling high burnup fuel.

The staffs review of the input and depletion parameters related to generating the source terms is discussed in this SER. The applicant uses the MCNP-5 version 1.40 code to calculate dose rates. MCNP is a transport theory based three-dimensional code that employs the Monte Carlo solution method. This code has been widely used across a wide range of applications and is well benchmarked and tested.

Response Functions The applicant generated response functions as discussed in section 5.4.1.1 of the application.

The response functions give the dose rate for per starting particle for each of the zones within the loading patterns neutron and for each gamma group and neutrons and secondary gammas.

The applicant provided the response functions in tables 5-53 through 5-64 of the application.

To benchmark the response function approach, the applicant stated that it chose four sample source terms and compared these to the dose rates predicted using MCNP directly.

The staff did not verify if these response functions have been generated accurately, however it performed independent calculations using UNF-ST&DARDS which calculates dose rates for a fully loaded system using a sampling of assemblies from the FQT. The staff found that using a different method than the applicant for calculating dose rates it was still able to confirm that the package meets regulatory requirements and therefore found that it has reasonable assurance that the FQTs generated by the applicant using this response function method are acceptable for this package.

Flux-to-Dose-Rate Conversion The applicant states in section 5.4.3 and 5.6.1.4.3 of the application that it uses the ANSI/ANS 6.1.1-1977 flux-to-dose rate conversion factors in all the shielding evaluations. The applicant provides the conversion factors in table 5-76 of the application for both gamma and neutron doses. The staff found this acceptable based on the review guidance in NUREG-2216.

Tallies To demonstrate that the package design meets the external dose rates at locations prescribed for the NCT vehicle surface, 2 meters, and HAC, the applicant is required to determine the maximum dose rate considering all points on the surface. The heterogeneous zone loading results in azimuthal variations in dose rate. The axial burnup profile and the presence of the shielding rings causes axial variations in dose rate. The applicant must use sufficiently small tally bins such that a maximum is not reduced to an average considering these features. The applicant shows a general representation of the tally locations for the TN Eagle LC with the EOS-37PTH basket in Figure 5-18 of the application. This figure shows that tally locations are chosen at multiple locations axially and at 2 meters.

The applicant discussed the tallies that it used in the response function method in section 5.4.1.1 of the application. The tallies are discussed in section 5.4.4.1 of the application under NCT and section 5.4.4.2 of the application under HAC. The staff reviewed the tally locations and

63 size of the tallies and found that it considered dose rates around every location of the package and at all regulatory locations of interest (surface, 2 meters under NCT and 1 meter under HAC).

5.5 Staff Calculations Source Term Calculations The staff used the ORIGAMI code from the SCALE 6.2.3 code package to verify the spent fuel source term for the new loading patterns and fuel qualification strategy. The staff used the bounding ARP libraries from the UNF-ST&DARDS code which were determined to be bounding with respect to the neutron source (Banerjee, Robb, Radulescu, Scaglione, Wagner, Clarity, LeFebvre, Peterson, Estimation of Inherent Safety Margins in Loaded Commercial Spent Nuclear Fuel Casks, Nuclear Technology, March 27, 2017, https://doi.org/10.13182/NT15-112; Radulescu, Banerjee, LeFebvre, Miller & Scaglione, Shielding Analysis Capability of UNF-ST&DARDS, Nuclear Technology, September 1, 2017, https://doi.org/10.1080/00295450.2017.1307643).

The staff independently calculated the spent fuel gamma source terms from tables 5-9 through 5-21 of the application. The source term determined by the staff was generally higher than the applicants source terms. For the lower energy groups the difference was less pronounced (10 percent or less) but some energy groups differed significantly and for higher energy groups the difference was more pronounced (greater than 20 percent). Lower energy groups have a much higher activity than the higher energy groups but are also more likely to get attenuated by shielding.

To determine how individual group differences would affect overall dose, the staff used Microshield 12.11x to compare the dose rates of the source term from staffs calculations to that of the source terms from tables 5-9 through 5-21 of the application by shielding the source through a representative amount of uranium and steel to account for assembly self-shielding and package shielding. The staffs calculations generally show that dose rate using the source term from the staffs calculations are higher (10-30 percent) than that from the application and can be significantly more so for BWR fuel (100+ percent).

The staff did not seek to determine the differences between the applicants depletion parameters and those used in its calculations because the staffs calculations using UNF-ST&DARDS (discussed in this SER) show that the package meets regulatory dose rate requirements. The staff expects that these differences are likely due to the difference in assumptions related to moderator density, and/or depleting fuel with burnable absorbers or control blades present.

UNF-ST&DARDS uses a very fine energy group structure. The staff performed a sensitivity study comparing the UNF-ST&DARDS results using the applicants energy group structure to the default used in UNF-ST&DARDS and found that the applicants group structure produces more conservative results. The staff found that assuming a coarse group structure is conservative when evaluating dose rate.

Even though the staff was unable to verify the applicants source term, the staffs overall dose rate calculations give the staff confidence that there is either (1) enough margin to the regulatory limit to compensate for uncertainties related to the source term determination, and/or (2) there are other conservatisms related to the shielding evaluation as a whole (such as grouping the source term is a coarse structure, performing a more detailed 3D Monte Carlo simulation to

64 account for transport and shielding, contributions from the neutron source, etc.) that are able to compensate for these uncertainties in gamma source term generation.

The staff did compare its overall neutron source to that of the applicants and it was slightly higher, but in relative agreement. Because the applicant does not use the neutron source from the ORIGAMI calculations, staff did not perform an energy group by group comparison.

Shielding Calculations The staff performed confirmatory calculations using UNF-ST&DARDS code Version 4.0 available through the Radiation Safety Information Computational Center (RSICC) at ORNL.

UNF-ST&ARDS is a comprehensive integrated data and analysis tool being developed for the DOE Office of Nuclear Energy (NE) Spent Fuel and Waste Disposition (SFWD) program with support from the NRC. UNF-ST&DARDS simplifies and automates performance of spent fuel analyses.

This code uses ORIGAMI for source term evaluations and Monaco/MAVRIC for the dose calculations. The staffs model was built by ORNL using design data from the FSAR and FSAR drawings. Some of the notable differences between the staffs calculation method and that of the applicant is that the staffs model represents the fuel rods/pins explicitly versus using a homogenized fuel mixture. The burnup and void fraction (BWR fuel) profiles are also represented by depleting each axial zone individually rather than using an adjustment factor.

The staffs model does not employ the response function method and includes the full loading pattern with all fuel assemblies modeled simultaneously using the design basis burnup, enrichment, and cooling time (BECT) combinations.

The staffs model also uses the neutron source from the ORIGAMI depletion and accounts for sub-critical multiplication explicitly within using the neutron transport models within Monaco/MAVRIC. The staffs model is also capable of incorporating failed fuel directly into the models.

The staff calculated the dose rate for following configurations under NCT:

65 Basket &

Cask Variant HLZC and Plan Source Term (BECT and CCs)

Design Basis from application Table 5-71; CCs in Zones 1 and 2 Design Basis from application Table 5-71; CCs in all zones Design Basis from application Table 5-71; No CCs; Failed fuel per Figure 1.6.1-1 of the application; 37PTH LC HLZC1 Plan1; 492kg U per FA Burnup and enrichment selected from application Tables 8.7.1-2 through 4 to correspond to a cooling time of about 4.6 years; failed fuel per Figure 1.6.1-1 of the application; CCs in all other locations Design Basis from application Table 5-71; No CCs Design Basis from application Table 5-71; CCs in Zones 1 and 2 37PTH LC HLZC1 Plan2; 492kg U per FA Design Basis from application Table 5-71; CCs in all zones Design Basis from application Table 5-71; CCs in Zones 1, 2, 3, 5 (inner zones)

Burnup and enrichment selected from application Tables 8.7.1-8 & 9 and 8.7.1-14 through 17 to correspond to a cooling time of about 5 years; CCs in Zones 1, 2, 3, 5 (inner zones)

Burnup and enrichment selected from application Tables 8.7.1-8 & 9 and 8.7.1-14 through 17 to correspond to a cooling time of about 5 years; Cobalt increased by 800 ppm of reconstituted fuel mass from application Table 5-33 in all zones, Zone 4 and 6 increased cooling time by 2 years; CCs in Zones 1, 2, 3, 5 (inner zones)

Design Basis from application Table 5-71; CCs in all zones 37PTH LC HLZC2 Plan2; 492kg U per FA Burnup and enrichment selected from application Tables 8.7.1-8 & 9 and 8.7.1-14 through 17 to correspond to a cooling time of about 5 years; CCs in all zones Design Basis from application Table 5-71 Design Basis from application Table 5-71; reduced fuel hardware Co-60 to correspond to applicants assumptions Burnup and enrichment selected from application Tables 8.7.2-2 through 4 to correspond to a cooling time of about 5 years 89BTH LC 198kg U per FA Burnup and enrichment selected from application Tables 8.7.2-2 through 4 to correspond to a cooling time of about 5 years; Cobalt increased by 800 ppm of reconstituted fuel mass from application Table 5-34 in all zones, Zone 3 increased cooling time by 6 years DBS4 per application Table 5.6.1-6 32PTH1L SC Type B 490kg U per FA; CC in all locations DBS1 per application Table 5.6.1-3 DBS4 per application Table 5.6.1-6 DBS1 per application Table 5.6.1-3 32PTH1L SC Type C 490kg U per FA; CC in all locations DBS1 per application Table 5.6.1-3; Cobalt increased by 800ppm of 2/3 reconstituted fuel mass from application Table 5-33 in all zones The staff has drawn the following observations/conclusions from its analyses:

66 The 2-meter radial dose rate is the most limiting for all TN Eagle variants.

The bottom location is most often the location for the limiting surface dose rate. This did not agree with the applicants evaluations. The staff did not investigate the discrepancy between its calculations and the applicants given the fact that there was so much margin between surface dose rates and the regulatory limit in 10 CFR 71.47(b). Also due to there being a hole in the center of the impact limiter that brings the tally surface closer to the package, higher dose rates at this location are less of a safety concern as personnel would not be inside this hole and the localized increase from the impact limiter design causes the dose rate at 2 meters to be significantly less than the dose rate 2 meters from the radial side.

89BTH LC is the most limiting design in the TN Eagle family of casks, this may be due to the large differences in source term between the staffs calculations and the applicants calculations.

32PTH1 has lower dose rates than the EOS baskets. This is expected as there are fewer assemblies. The staff determined that it had confidence that this basket bounds the other baskets allowed in the SC cask variant given that they all use the same FQT.

The staff calculations showed that some loading patterns with lower cooling times were more limiting than the design basis loadings that the applicant used from table 5-71 and DBS4 (table 5.6.1-6) in the application. As discussed in this SER, this is likely due to the applicants assumptions that would result in lower Co-60. Additional Co-60 would have a more prominent effect on dose rates for loading with lower cooling times as the half-life of Co-60 is around 5 years. Therefore, differences between the staffs model and the applicants model would also be more prominent for loadings with lower cooling times.

All PWR variants of the cask meet regulatory dose rate limits. Some of the staffs calculations for BWR loadings resulted in calculated dose rates above regulatory limits for the 2-meter dose rate (less than 7 percent). This is likely due to differences in the source term but also due to larger amount of Co-60 assumed in the staffs model. When using the same assumptions related to Co-59 impurities as the applicant, the staffs calculations show that the TN Eagle with the 89BTH meets regulatory limits even with the discrepancies in the source terms.

The additional cooling time associated with reconstituted fuel assemblies is enough to account for additional Co-60 from activated stainless steel replacement rods. There is no additional cooling time required for the DSCs allowed in the TN Eagle-SC, however, staff calculations show that there is enough margin to the regulatory limit to account for the increase in dose rate due to the presence of activated stainless steel rods.

Similar to the applicants model, under HAC, the UNF-ST&DARDS model created by ORNL has tallies adjusted to account for the loss of the impact limiter and removes the two central neutron shield rings. It also reduced the resin density accordingly to account for the loss of material from the HAC fire. The staff had the capability of modeling failed or intact fuel. It performed the following calculations:

67 Basket Loading Pattern Source Term 37PTH LC HLZC1 Plan1; 492kg U per FA; all failed; CCs in Zones 1 and 2 Design Basis from application Table 5-71 32PTH1L SC Type B

490kg U per FA; CC in all locations; Intact fuel DBS4 per application Table 5.6.1-6 89BTH LC 198kg U per FA; Intact Fuel Design Basis from application Table 5-71 89BTH LC 198kg U per FA; all fuel failed Design Basis from application Table 5-71 All of the calculations run by the staff for HAC showed significant margin to the regulatory limit in 10 CFR 71.51(a)(2).

Overall given the different approaches to calculate dose rates by the applicant versus the staff and the vast amount of cask variants and contents, everything modeled by the staff resulted in dose rates that met (or slightly exceeded, see above on cobalt assumptions) the regulatory limit.

This gives the staff confidence that the applicant has appropriately modeled the TN Eagle and that as specified, meets regulatory dose rate limits in 10 CFR 71.47(b) and 71.51(a)(2).

5.6 Evaluation Findings

The staff has reviewed the application and finds that it adequately describes the package contents, and the package design features that affect shielding in compliance with 10 CFR 71.31(a)(1), 71.33(a), and 71.33(b), and provides an evaluation of the packages shielding performance in compliance with 10 CFR 71.31(a)(2), 71.31(b), 71.35(a), and 71.41(a). The descriptions of the packaging and the contents are adequate to allow for evaluation of the packages shielding performance. The evaluation is appropriate and bounding for the packaging and the package contents as described in the application.

The staff has reviewed the application and finds that it demonstrates the package has been designed so that under the evaluations specified in 10 CFR 71.71 (normal conditions of transport), and in compliance with 10 CFR 71.43(f) and 10 CFR 71.51(a)(1), the external radiation levels do not significantly increase.

The staff has reviewed the application and finds that it demonstrates that under the evaluations specified in 10 CFR 71.71 (normal conditions of transport), external radiation levels do not exceed the limits in 10 CFR 71.47(a) for nonexclusive-use shipments or 10 CFR 71.47(b) for exclusive-use shipments, as applicable.

The staff has reviewed the application and finds that it demonstrates that under the tests specified in 10 CFR 71.73 (hypothetical accident conditions), external radiation levels do not exceed the limits in 10 CFR 71.51(a)(2).

The staff has reviewed the application and finds that it identifies codes and standards used in the packages shielding design and in the shielding analyses, in compliance with 10 CFR 71.31(c).

The staff has reviewed the application and finds that it includes operations descriptions, acceptance tests, and maintenance programs that will ensure that the package is fabricated,

68 operated, and maintained in a manner consistent with the applicable shielding requirements of 10 CFR Part 71.

Based on its review of the information and representations provided in the application and the staffs independent, confirmatory calculations, the staff has reasonable assurance that the proposed package design and contents satisfy the shielding requirements, and the radiation level limits in 10 CFR Part 71. The staff also considered the regulation itself, appropriate regulatory guides, applicable codes, and standards, and accepted engineering practices, in reaching this finding.

6.0 CRITICALITY EVALUATION

This section presents the findings of the criticality safety review for CoC No. 9382 for the Model No. TN Eagle transportation package. This CoC application includes a criticality analysis that credits reduced reactivity due to fuel burnup for some contents. The staff evaluated the package for its ability to meet the fissile material requirements of 10 CFR Part 71, including the general requirements for fissile material packages in 10 CFR 71.55, and the standards for arrays of fissile material packages in 10 CFR 71.59. The staff reviewed the criticality safety analysis of the package presented in the SAR, and also performed independent calculations to confirm the applicants results. The staffs review considered the criticality safety requirements of the radioactive material transportation regulations in 10 CFR Part 71, as well as the review guidance presented in NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material.

6.1 Description of the Criticality Design 6.1.1 Packaging Design Features The Model No. TN Eagle design consists of a cylindrical, steel shell containment system, with a flat bottom and bolted closure lid at the top, and with seven different available DSCs with internal basket structures for maintaining the position of the spent fuel contents. Each DSC type consists of a stainless-steel cylinder with a double seal-welded closure, and with similar basket structural materials and geometry, as described in the SAR. Criticality safety is maintained by the fixed geometry of the basket, as well as by the borated aluminum, aluminum/B4C metal matrix composite, or Boral neutron poison plates present between each spent fuel assembly location.

The applicant requested the TN Eagle package be authorized to transport the EOS 89BTH, EOS 37PTH, FO-FC-FF, 24PT1, 32PTH1, 24PT4, and 32PT DSCs. The FO-FC-FF and 24PT1 DSCs were previously evaluated and approved for transportation in the TN NUHOMS MP-187 Multi-Purpose Cask (Certificate of Compliance No. USA/9255/B(U)F-85). The 32PTH1, 24PT4, and 32 PT DSCs were previously evaluated and approved for transportation in the TN NUHOMS-MP197HB package (Certificate of Compliance No. USA/9302/B(U)F-96).

6.1.2 Codes and Standards The applicable regulations considered in the review of the criticality safety portion of this application include the fissile material requirements in 10 CFR Part 71, specifically the general requirements for fissile material packages in 10 CFR 71.55, and the standards for arrays of fissile material packages in 10 CFR 71.59. The staff also used the review guidance contained in NUREG-2216.

69 6.1.3 Summary Table of Criticality Evaluations The applicant provided a summary of maximum calculated keffs for all transportation conditions in section 6.1.2 of the SAR. The maximums reported under any condition are summarized in the following table for all DSCs. All results include the calculated keff plus two times the Monte Carlo uncertainty, and for DSCs evaluated using burnup credit, the bias due to isotopic depletion and minor actinide and fission product criticality. The Upper Safety Limit (USL) includes the criticality code bias and bias uncertainty determined in the benchmarking analysis.

DSC Keff USL EOS 89BTH 0.9398 0.9418 EOS 37PTH 0.9413 0.9423 FO-FC-FF 0.9497 0.9500 24PT1 0.9392 0.9500 24PT4 0.9393 0.9411 32PTH1 0.9375 0.9412 32PT 0.9419 0.9423 6.1.4 Criticality Safety Index The applicant demonstrated that infinite arrays of TN Eagle packages are subcritical under normal conditions of transport and hypothetical accident conditions. Therefore, the criticality safety index (CSI), determined in accordance with 10 CFR 71.59(b), is 0.0.

6.2 Spent Nuclear Fuel Contents The TN Eagle package is designed to transport a maximum of 89 BWR and 37 PWR fuel assemblies in DSCs. Specific DSC contents are summarized below:

EOS 89BTH - designed to transport up to 89 intact or reconstituted 7x7, 8x8, 9x9, or 10x10 BWR spent fuel assemblies with or without channels. Fuel assembly specifications are contained in table 1.6.2-2. Fuel assembly maximum lattice average initial enrichment and neutron poison requirements are given in table 1.6.2-3.

EOS 37PTH - designed to transport up to 37 intact or reconstituted PWR spent fuel assemblies, with or without control components, up to 8 damaged PWR fuel assemblies with the remaining fuel intact, or up to 4 failed PWR fuel assemblies with the remaining fuel intact. Fuel assembly specifications are given in table 1.6.1-15. Initial enrichment, burnup, and cooling time requirements determined in the burnup credit analysis are given in the loading curve for intact and damaged fuel configurations in tables 1.6.1-2 through 1.6.1-13.

NUHOMS-24PT4 - designed to transport up to 24 intact or reconstituted Combustion Engineering 16x16 PWR spent fuel assemblies, or up to 12 damaged assemblies in failed fuel cans, with the remaining fuel intact. Fuel assembly design specifications are contained in table 1.6.3-3. Fuel assembly maximum initial enrichment and neutron poison requirements are given in table 1.6.3-4.

NUHOMS-32PT - designed to transport up to 32 intact or reconstituted PWR spent fuel assemblies, with or without control components. Fuel assembly specifications are given in table 1.6.4-3. Initial enrichment, burnup, cooling time and poison rod assembly requirements determined in the burnup credit analysis are given for all fuel assembly classes except

70 CE14x14 in the loading curve in table 1.6.4-6. Initial enrichment, burnup, and cooling time requirements determined in the burnup credit analysis are given for intact CE14x14 class fuel assemblies in table 1.6.4-7, for damaged CE14x14 class fuel assemblies in table 1.6.4-8, and for failed CE14x14 class fuel assemblies in table 1.6.4-9. Poison rod assembly requirements for loading are given in table 1.6.4-5.

NUHOMS-32PTH1 - designed to transport up to 32 intact or reconstituted PWR spent fuel assemblies, with or without control components, or up to 16 damaged fuel assemblies with the remaining fuel intact. Fuel assembly specifications are given in table 1.6.5-4. Initial enrichment, burnup, cooling time, and neutron absorber type requirements determined in the burnup credit analysis are given in the loading curves for intact and damaged fuel configurations in tables 1.6.5-7 and 1.6.5-8, respectively.

NUHOMS-FO-FC-FF DSCs - designed to transport up to 24 intact or reconstituted B&W 15x15 PWR spent fuel assemblies in the FO or FC DSCs, or up to 13 damaged or failed B&W 15x15 fuel assemblies in failed fuel cans in the FF DSC. The maximum enrichment for fuel transported in these DSCs is 3.43 weight percent U-235. Fuel assembly design specifications are contained in table 1.6.6-2.

NUHOMS-24PT1 - designed to transport up to 24 intact WE 14x14 stainless steel clad UO2 PWR spent fuel assemblies, or up to 24 intact WE 14x14 zirconium alloy clad mixed oxide (MOX) damaged or failed fuel assemblies. This DSC may also transport up to 4 damaged or failed UO2 fuel assemblies, or 1 MOX fuel assembly, in failed fuel cans, with the remaining fuel intact. Fuel assembly specifications, including maximum initial enrichment for UO2 fuel assemblies and uranium and plutonium contents for MOX fuel assemblies, are given in table 1.6.7-1. Fuel assembly characteristics important for criticality safety are given in table 1.6.7-2.

6.3 General Considerations for Criticality Evaluations The FO-FC-FF and 24PT1 DSCs were previously evaluated and approved for transportation in the TN NUHOMS MP-187 Multi-Purpose Cask (Certificate of Compliance No.

USA/9255/B(U)F-85). The requested contents for these DSCs transported in the TN Eagle package are identical to those previously evaluated and approved for the NUHOMS MP-187 Multi-Purpose Cask. Since these DSCs and their contents are identical to that previously evaluated, and the TN Eagle package materials and dimensions are like those for the NUHOMS MP-187 Multi-Purpose Cask, no additional criticality analyses are performed for the FO-FC-FF and 24PT1 DSCs in the TN Eagle package.

Similarly, the 32PTH1, 24PT4, and 32 PT DSCs were previously evaluated and approved for transportation in the TN NUHOMS-MP197HB package (Certificate of Compliance No.

USA/9302/B(U)F-96). The requested contents for these DSCs transported in the TN Eagle package are identical to those previously evaluated and approved for the NUHOMS-MP197HB package, except for CE 14x14 fuel in the 32PT DSC. Since these DSCs and their contents are identical to that previously evaluated (other than CE 14x14 fuel in the 32PT), and the TN Eagle package materials and dimensions are like those for the NUHOMS-MP197HB package, no additional criticality analyses are performed for the 32PTH1, 32PT (with contents other than CE 14x14 fuel), and 24PT4 DSCs in the TN Eagle package.

The staff agrees that the previous criticality analysis for the FO-FC-FF, 24PT1, 32PTH1, 24PT4, and 32PT DSCs, except for CE14x14 fuel in the 32PT, are applicable to the TN Eagle package.

71 All further discussion of the TN Eagle criticality evaluation in this SER will be for criticality analysis of the EOS 89BTH, EOS 37PTH, and 32PT (with CE 14x14 fuel) DSCs.

6.3.1 Model Configuration The applicant evaluated three-dimensional models of a single package and arrays of packages under both normal conditions of transport and hypothetical accident conditions. The applicant explicitly modeled the fuel rods and cladding, guide tubes, water gaps, basket structure, and neutron absorber in the basket. The criticality models for the TN Eagle package include assumptions intended to produce a conservative estimate of system keff. These assumptions include:

Modeling the fuel pellet stack at a minimum of 96.5 percent of theoretical uranium oxide density, with no allowance for pellet dishing or chamfer, Flooding of the fuel-pellet gap with full density water, Uniform lattice-average or assembly-average enrichments, with no credit for burnable absorbers or natural uranium blankets, Most reactive combination of material and fabrication tolerances, and eccentric positioning of fuel assemblies in the basket guide tubes, Neglecting package spacing provided by the impact limiters, For hypothetical accident conditions, replacing the external neutron shield and stainless-steel skin with water, and Neglecting fuel assembly structural material that would absorb neutrons or displace moderator, such as grid plates, spacers, rod plenums, end fittings, and channels above and below the active fuel.

Damaged fuel assemblies must be placed in basket locations with top and bottom end caps in DSC types for which damaged fuel is authorized. The applicant modeled damaged fuel in various configurations to find the most reactive condition.

6.3.2 Material Properties Fresh fuel compositions were modeled as UO2 with a minimum 96.5 percent of theoretical density. The U-234 and U-236 which are present in fresh fuel, are conservatively ignored. For burned fuel compositions, the applicant modeled the fuel with actinide and fission product nuclides incorporated into the fuel matrix, consistent with the recommendations in section 6.4.7 of NUREG-2216. Section 6.8.2.6 of the SAR discusses the calculations to determine burned fuel compositions for the EOS-37PTH DSC. Section 6.8.5.6 of the SAR discusses the calculations to determine burned fuel compositions for the 32PT DSC with CE 14x14 class spent fuel.

The individual SAR sections 6.8.1 through 6.8.5 provide the composition of the major components of each DSC, including fresh and burned UO2 fuel, steel and aluminum structural components, and neutron absorber panels, poison rodlets, and poison rod assemblies.

The criticality analyses conservatively assume no more than 75 percent of the neutron absorber manufacturers minimum specified B-10 content for Boral panels, as well as for B4C poison rodlets and poison rod assemblies. For borated aluminum and aluminum/B4C metal matrix composite neutron absorbers, the criticality analyses assume no more than 90 percent of the

72 minimum specified B-10 content, due to the more robust material verification requirements for these materials.

6.3.3 Analysis Methods and Nuclear Data For the criticality analysis of the EOS 89BTH DSC, the applicant used the CSAS5 sequence of the SCALE 6.0 computer code system, with the KENO V.a three-dimensional Monte Carlo neutron transport code and the 44-group ENDF/B-V cross-section library, as discussed in section 6.8.1.2.3 of the SAR. For the fresh fuel criticality analyses of the EOS 37PTH DSC and CE 14x14 fuel in the 32PT DSC, the applicant used the CSAS5 sequence of the SCALE 6.1.3 computer code system, with the KENO V.a three-dimensional Monte Carlo neutron transport code and the 238-group ENDF/B-VII cross section library.

For burnup credit calculations in the EOS 37PTH DSC and CE 14x14 fuel in the 32PT DSC, the applicant used the STARBUCS burnup credit criticality sequence of the SCALE 6.1.3 computer code system, with the ORIGEN-ARP isotopic depletion code, KENO V.a three-dimensional Monte Carlo neutron transport code, and the 238-group ENDF/B-VII cross section library. The SCALE code is a standard in the nuclear industry for performing Monte Carlo criticality safety and radiation shielding calculations.

6.3.4 Demonstration of Maximum Reactivity The applicant performed multiple sensitivity studies for the single package and array configurations of the TN Eagle to determine the most reactive condition. These studies included variation of internal and external moderation, partial flooding of the containment (when possible), eccentric positioning of fuel assemblies in the basket, and variation of basket structural and neutron absorbing materials according to the worst-case combination of material and fabrication tolerances.

The applicants sensitivity studies to determine the most reactive configuration are summarized in SAR sections 6.8.1.2.4, 6.8.2.4, and 6.8.5.6 for the EOS 89BTH, EOS 37 PTH, and 32PT DSCs, respectively.

6.3.5 Confirmatory Analyses The staff performed confirmatory criticality evaluations of the TN Eagle package, for both fresh and burned fuel configurations, using the ORIGAMI isotopic depletion sequence and the KENO VI three-dimensional Monte Carlo criticality sequence in the SCALE 6.2.4 code system, and the continuous energy ENDF/B-VII.1 cross section library.

Using assumptions like the applicants, the staff calculated keff values for select configurations which were within the margin of error of those calculated by the applicant and confirmed that the package will meet the criticality safety requirements of 10 CFR Part 71.

6.4 Single Package Evaluation 6.4.1 Configuration The applicant modeled a single TN Eagle package, with each DSC as described in detail in the SAR sections 6.8.1, 6.8.2, and 6.8.5, assuming full moderation by water, with fuel in the as-loaded condition to demonstrate subcriticality per the requirements of 10 CFR 71.55(b).

73 6.4.2 Results The results of the single package keff evaluations are contained in each of the SAR sections 6.8.1, 6.8.2, and 6.8.5. The highest keff results are for the as-loaded configuration under 10 CFR 71.55(b), all of which are below the applicants calculated USL.

6.5 Evaluation of Package Arrays 6.5.1 Package Arrays Under Normal Conditions of Transport The applicant modeled infinite arrays of TN Eagle packages under HAC. This analysis bounds the condition of package arrays under NCT, and therefore, the applicant shows that infinites arrays of TN Eagle packages are subcritical under both NCT and HAC.

6.5.2 Package Arrays Under Hypothetical Accident Conditions The applicant reflected the most reactive single package models in the criticality model on all sides to produce infinite arrays of packages, conservatively ignoring the impact limiter, neutron shield, and neutron shield stainless steel skin. The applicant varied the water density between packages to find the most reactive condition.

6.5.3 Package Arrays Results and CSI Infinite array keff values are not significantly different than those for the water-reflected single package models. This indicates that the package is isolated with respect to neutron interaction, as expected due to the large amount of structural and shielding material between fissile material in adjacent packages.

Infinite array keff results for each of the DSCs in the TN Eagle package are contained in the SAR sections 6.8.1, 6.8.2, and 6.8.5 for the EOS 89BTH, EOS 37PTH, and 32PT DSCs, respectively.

Since an infinite array of packages are demonstrated to be subcritical under both normal conditions of transport and hypothetical accident conditions, the resulting CSI is 0.0.

6.6 Fresh Fuel Benchmark Evaluations 6.6.1 Experiments and Applicability The applicant performed a fresh fuel benchmarking analysis for the EOS 89BTH DSC, and the fresh fuel analyses in the EOS 37PTH and 32PT DSCs, using fresh uranium oxide experiments, chosen to have characteristics similar to the TN Eagle package. The experiment parameters evaluated for range of applicability and trends included:

235U enrichment, Fuel rod pitch, Water/fuel volume ratio, Assembly separation, Hydrogen-to-fissile (H/X) ratio, Average fission energy group (AEG), and Energy of the average lethargy of fission (EALF)

74 The applicant evaluated each critical experiment using the same code, cross section library, computer platform, and modeling techniques as was used for the criticality evaluation.

6.6.2 Bias Determination The applicant determined the USL for the TN Eagle package using the single-sided lower tolerance limit recommended in NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology.3 The applicant determined a separate USL for fresh fuel calculations in each of the DSCs evaluated, summarized in sections 6.8.1.7.2, 6.8.2.4.2, and 6.8.5.6.2 for the EOS 89BTH, EOS 37PTH, and 32PT DSCs, respectively.

The trending parameter values from the limiting criticality analysis was compared to the USL equations for those parameters to determine a set of USLs based on each parameter. The most limiting (i.e., lowest) USL from this set was chosen as the USL for that DSC. The resulting USLs are summarized in the following table:

DSC Bounding USL EOS 89BTH 0.9418 EOS 37PTH 0.94173 32PT 0.94173 The staff reviewed the fresh fuel benchmarking analysis performed by the applicant and determined that the USL was determined in accordance with relevant NRC guidance. The critical experiments chosen are appropriate for the system being evaluated, and the resulting USLs are bounding and acceptable.

6.7 Burnup Credit Evaluation for Commercial Light-Water Reactor (LWR) Spent Nuclear Fuel (SNF) 6.7.1 Limits for the Certification Basis The TN Eagle package burnup credit criticality evaluations for the EOS 37PTH and 32PT DSCs assume a maximum of 5.0 weight-percent U-235 initial enrichment, credit a burnup of no more than 60 GWd/MTU, and take credit for cooling time of no more than 40 years, all of which meet the certification basis limits in NUREG-2216.

6.7.2 Model Assumptions 6.7.2.1 Isotopic Depletion For burned fuel criticality calculations for PWR fuel in the EOS 37PTH DSC and CE 14x14 fuel in the 32PT DSC, the applicant used the STARBUCS sequence of the SCALE 6.1.3 code system. This sequence uses reactor libraries with isotopic compositions as a function of burnup, enrichment, and cool time for a particular fuel type, pre-generated with the TRITON two-dimensional isotopic depletion sequence in the SCALE 6.1.3 code.

The applicant generated specific reactor libraries for all fuel assemblies modeled in the burnup credit criticality analysis, assuming that either a BPRA was present during the entire irradiation, or that control rods were fully inserted during the first 15 GWd/MTU of irradiation in the reactor.

The applicant stated that these absorber exposures were bounding, as BPRAs are rarely

75 inserted for more than one cycle of irradiation, and control rods are rarely inserted beyond the top of the fuel assembly for any amount of irradiation.

For PWR fuel in the EOS 37PTH DSC, the applicant determined that the TN Eagle system was more reactive with fuel that had been exposed to BPRAs for the entire fuel irradiation. For CE 14x14 fuel, BPRAs are not used during irradiation, so the applicant modeled this fuel in the 32PT DSC using the reactor libraries generated with full control rod insertion for the first 15 GWd/MTU of irradiation. The staff agrees with the applicants conclusion that the absorber exposures assumed in the generation of reactor libraries for use in the STARBUCS criticality calculations are conservative, since they bound absorber exposures anticipated for any fuel type during irradiation in a reactor.

For the burnup credit criticality calculations in the TN Eagle package, the applicant considered the effects of axial burnup distribution. The applicant assumed a burnup-dependent, 18-node axial profile, as shown in table 6.8.2-6 of the SAR. The staff compared the applicants axial profiles to the conservative profiles determined in NUREG/CR-6801, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses,2 and determined that the profiles are consistent with the profiles recommended in this NUREG/CR.

The applicant estimated the isotopic composition of burned fuel as a function of burnup, enrichment, and cooling time, which was subsequently used in the criticality calculations. The model assumptions for the isotopic depletion analysis that affect the reactivity of the package include the core specific power, moderator temperature, fuel temperature, and soluble boron concentration during irradiation. These core operating parameters are given in table 6.8.2-30 of the SAR for the EOS 37PTH DSC, and in table 6.8.5-8 of the SAR for the 32PT DSC. These parameters are also included as loading criteria in section 8.4.1 of the package Operating Procedures.

6.7.2.2 Criticality For the criticality calculations in DSCs evaluated using burnup credit, the applicant considered the effects of axial burnup distribution. The STARBUCS sequence has built-in profiles that are based on the limiting axial profiles from NUREG/CR-6801. The applicant used these profiles in the criticality analysis, as well as one for higher burnup (>38 GWd/MTU) based on NUREG/CR-6801.

The criticality model for burnup credit includes all the limiting features determined for the fresh fuel analyses and scoping calculations. These include a minimum fuel pellet stack density of 96.5 percent of theoretical, no allowance for pellet dishing or chamfer, flooding of the fuel-pellet gap with full density water, no credit for burnable absorbers or natural uranium blankets, most reactive combination of material and fabrication tolerances, eccentric positioning of fuel assemblies in the basket guide tubes, neglecting package spacing provided by the impact limiters, replacing the external neutron shield and stainless steel skin with water, and neglecting fuel assembly structural material that would absorb neutrons or displace moderator, such as grid plates, spacers, rod plenums, end fittings, and channels above and below the active fuel.

6.7.3 Code Validation - Isotopic Depletion NUREG-2216 recommends that applicants may use pre-calculated values for isotopic depletion code bias and bias uncertainty, provided the applicant uses the SCALE code system with TRITON and the 238-group ENDF/B-VII cross section library, and that the evaluated system is neutronically similar to the GBC-32 system evaluation in NUREG/CR-7108, An Approach for

76 Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses - Isotopic Composition Predictions.4 The applicant used the code and cross section library referenced in NUREG-2216, and provided information that demonstrates that the TN Eagle system is similar to the GBC-32 in section 6.8.2.7 of the SAR.

6.7.4 Code Validation - keff Determination The applicant modeled several critical experiments, chosen to have characteristics similar to the TN Eagle package with burned fuel contents. The selected experiments are described in section 6.8.2.5.2 of the SAR and listed in tables 6.8.2-19 and 6.8.2-20 of the SAR. The experiment parameters evaluated for range of applicability and trends included:

235U enrichment, Fuel rod pitch, Water/fuel volume ratio, Assembly separation, H/X ratio, AEG, and EALF The applicant evaluated each critical experiment using the same code, cross section library, computer platform, and modeling techniques as was used for the criticality evaluation.

The applicant determined the USL for the major actinides in the TN Eagle package using the single-sided lower tolerance limit recommended in NUREG/CR-6698. The applicant determined a separate USL for burnup credit calculations in each of the DSCs evaluated, summarized in sections 6.8.2.5.3 and 6.8.5.6.2 for the EOS 37PTH and 32PT DSCs, respectively.

The trending parameter values from the limiting criticality analysis was compared to the USL equations for those parameters to determine a set of USLs based on each parameter. The most limiting (i.e., lowest) USL from this set was chosen as the USL for that DSC. The resulting limiting USL chosen by the applicant to bound both the EOS 37PTH and 32PT DSC burnup credit criticality analyses is 0.94236.

For benchmarking the minor actinide and fission product component of the burnup credit analysis, the applicant applied 1.5 percent of the minor actinide and fission product worth to bound the bias and bias uncertainty from those nuclides, consistent with the recommendations of section 6.4.7.4 of NUREG-2216. Determination of minor actinide and fission product worth, and the resulting component of the USL, is discussed in section 6.8.2.5.6 of the SAR.

The staff reviewed the burnup credit criticality benchmarking analysis performed by the applicant and determined that the USL was determined in accordance with relevant NRC guidance. The critical experiments chosen are appropriate for the system being evaluated, and the resulting USLs are bounding and acceptable.

6.7.5 Loading Curve and Burnup Verification The loading curves for the DSCs previously evaluated and approved for transport in the NUHOMS-MP197HB package are repeated for the 32PTH1 and 32PT DSCs (except for CE

77 14x14 class fuel) in sections 1.6.5 and 1.6.4, respectively. These loading curves are identical to those associated with transportation of these DSCs in the NUHOMS-MP197HB.

The loading curves for DSCs evaluated using burnup credit in the TN Eagle package are given in sections 1.6.1 and 1.6.4 of the SAR for the EOS 37PTH and 32PT DSCs, respectively. The loading curves for the EOS 37PTH DSC in section 1.6.1 of the SAR include three sets of curves: one for B&W 15x15 class fuel assemblies, one for WE 17x17, WE 15x15, CE 14x14, CE 15x15, and CE 16x16 class fuel assemblies, and one for WE 14x14 class fuel assemblies. Each set includes curves for intact or damaged fuel, Type A or Type B basket neutron absorber areal density, and for 5, 10, 15, and 20 years of cooling time.

The loading curves for the 32PT DSC in section 1.6.4 of the SAR includes a single set for CE 14x14 class fuel assemblies, including curves for intact, damaged, and failed fuel, and for cooling times of 5, 10, 15, and 20 years.

NUREG-2216 includes an option that a misload analysis combined with additional administrative procedures for loading under burnup credit may be used in lieu of the recommended direct burnup measurement. This option is used in the TN Eagle analysis, with the misload analysis and additional administrative procedures summarized below.

6.7.5.1 Misload Analyses The applicant performed a comprehensive misload analysis demonstrating that the EOS 37PTH DSC with a single severely underburned fuel assembly, or multiple moderately underburned fuel assemblies, will still be adequately subcritical with freshwater intrusion. The applicant determined that all misloads considered would be subcritical with a reduced administrative margin of 0.02. The misload scenarios considered by the applicant are consistent with the recommendations in section 6.4.7.5 of NUREG-2216.

Misload analyses were also performed for the 32PTH1 and 32PT DSCs in the criticality analysis supporting transportation of these DSCs in the NUHOMS-MP197HB package. These misload analysis are applicable to the TN Eagle package, as the DSCs are identical, and the TN Eagle package materials and dimensions are like those of the NUHOMS-MP197HB package.

6.7.5.2 Administrative Procedures The applicant includes additional administrative procedures for loading DSCs evaluated for criticality safety using burnup credit. These procedures include:

A requirement for no fresh fuel in pool at time of loading, or verification that fuel being loaded into the canister is not fresh, either visually or by qualitative measurement.

A full pool audit within 1 year prior to loading, including visual verification of assembly identification numbers.

Identification of highly underburned and high reactivity fuel assemblies in the pool both prior to and after loading. Alternatively, the licensee can perform a misload evaluation to identify these highly underburned and high reactivity fuel assemblies. This evaluation will be subject to NRC review and approval.

A minimum required soluble boron concentration of 1800 ppm (based on minimum value from the 10 CFR Part 72 CoC No. 1004 UFSAR) in the pool for both loading and unloading.

78 A requirement that assemblies without visible identification number must have a quantitative confirmatory measurement prior to loading.

These additional procedures are comparable to those recommended in NUREG-2216 and are acceptable for reducing the likelihood and severity of misload events.

6.8 Findings

The staff finds the applicant has demonstrated that the TN Eagle package, when loaded with fuel assemblies meeting the characteristics of the contents described in appendices 1.6.1 through 1.6.7 of the SAR, will be adequately subcritical under all conditions. Therefore, the applicant has shown, and the staff finds that, the TN Eagle package meets the fissile material requirements of 10 CFR 71.55 for single packages, and 10 CFR 71.59 for arrays of packages with a CSI of 0.0 for PWR and BWR DSCs.

6.9 References 1.

U.S. Nuclear Regulatory Commission, Standard Review Plan for Transportation Package for Spent Fuel and Radioactive Material, NUREG-2216, August 2020.

2.

J.C. Wagner, M.D. DeHart, and C.V. Parks, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses, NUREG/CR-6801, (ORNL/TM-2001/273), U.S.

Nuclear Regulatory Commission, Oak Ridge National Laboratory, March 2003.

3.

J.D. Dean and R.W. Tayloe, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, U.S. Nuclear Regulatory Commission, Science Applications International Corporation, January 2001.

4.

G. Radulescu, I.C. Gauld, G. Ilas, and J.C. Wagner, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses - Isotopic Composition Predictions, NUREG/CR-7108, (ORNL/TM-2011/509), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, April 2012.

7.0 OPERATING PROCEDURES The package operations for spent fuel contents are based on those of the DSCs. The operations in chapter 8 of the application include procedures for loading a DSC from a horizontal fuel module (HFM) into the TN Eagle. The procedures also include steps to install the DSC spacers used to restrict axial movement and control the center of gravity, if needed.

Preparation of the TN Eagle to transport fuel offsite involves the following: (1) preparation of the cask for use, (2) verification that the fuel assemblies loaded in the DSC meet the required criteria, (3) loading of a DSC into the cask, (4) preparation of the cask for transport, (5) assembly verification leakage-rate testing of the packaging containment boundary, (6) placement of the cask onto a transportation vehicle, (7) installation of the impact limiters, and (8) performing final surveys and shipping paperwork.

Unloading of the TN Eagle after transport involves the following: (1) performing receipt inspections and surveys, (2) removing the impact limiters, (3) removing the package from the transportation vehicle, (4) unloading of the DSC into an HSM, and (5) preparing the packaging for shipment.

Preparing the packaging for shipment includes (1) Verification that the cask is empty to the extent possible, determination of the amount and form of residual internal activity within the

79 interior of the empty cask, (2) closing the empty cask (seals can be reused if found to be free of damage), (3) installing the primary lid and torque the bolts to between 20 percent and 50 percent of the maximum value specified in the applicable drawings and following the torque pattern shown in Figure 8-1, (4) installing the ram access cover plate and torque the bolts to at least 20 percent of the maximum value specified in the applicable drawings referenced in the CoC for package approval following the torque pattern shown in Figure 8-1, (5) installing the lid port plug and the lid port plug tightening nut, and (6) installing the lid port cover plate, the test port plugs into the primary lid and ram access cover plate.

The procedures for preparing the cask for transport include also a more detailed radiation survey when high burnup fuel is being shipped because such more detailed measurements may provide some indication if the fuel has failed as they are also required at package receipt and compare against the measurements taken before transport.

Staff requested the applicant to include the transportation frame in the numerical model of the TN Eagle package for drop analyses conducted under hypothetical accident conditions (HAC) to obtain assurance that the response of the impact limiters remains uninfluenced by the presence of the transportation frame. The current analysis did not include the potential influence of the frame on the response of the package: as such, the frame may impact the rigid surface prior to the engagement of the impact and if the frame is considered, the deformation of the impact limiters may be constrained. The applicant did not respond at this time and the certificate of compliance includes a condition to address that concern.

To ensure the DSCs, previously loaded under a 10 CFR Part 72 program, are acceptable for transportation, several verifications shall be performed prior to shipment. In particular, as already stated in this SER, there are additional administrative procedures for loading DSCs evaluated for criticality safety using burnup credit. There must also be a verification that the fuel assemblies to be transported in any DSC meet the characteristics, maximum average initial enrichment and burnup combinations for both intact and failed fuel assemblies as stated in appendix 1.6 of the application and are loaded per the fuel qualification tables in appendix 8.7 of the application. Finally, the aging management plan and evaluation for each DSC, or set of DSCs, shall be submitted to the NRC prior to shipment.

The staff reviewed the operating procedures in chapter 8 of the application and found that they contain the necessary operations for ensuring that contamination and regulatory dose rate limits will be met. Therefore, the staff determined that the procedures meet the requirement in 10 CFR 71.0(d)(3) which requires operating control requirements to measure radiation and contamination levels as stated in 10 CFR 71.87(i) and (j).

Appendix 8.7 of the application also specifies the allowable loading tables for the TN Eagle contents. The staff reviewed this information and determined that the allowable contents as specified are consistent with the assumptions used within the shielding analyses, as discussed in this SER.

The NRC staff has reviewed the description of the operating procedures and finds that the package will be prepared, loaded, transported, received, and unloaded in a manner consistent with its design. The NRC staff has reviewed the description of the special instructions to inspect, handle, and to safely open a package and concludes that the procedures for providing the special instructions to the consignee are in accordance with the requirements of 10 CFR 71.89.

8.0 ACCEPTANCE TESTS AND MAINTENANCE

80 Acceptance tests are mainly performed at the fabricators facility prior to delivery of the cask for use. These tests and inspections, which are only applicable to cask components identified as quality category A, B, or C on the applicable drawings for package approval, shall be performed in accordance with written procedures with the results being documented and retained.

Visual inspections shall be performed on all cask components for any evidence of damage or deformation such as, but not limited to, cracks, pinholes, and uncontrolled voids. Surface finish inspections shall be performed on all containment boundary sealing surfaces for conformance with the applicable drawings for package approval. Dimensional inspections shall be performed of the cask components to verify they conform to the applicable drawings for package approval.

Weld examinations should verify that the welds were performed using processes and personnel, both qualified in accordance with the applicable sections of the ASME B&PV Code; Dimensional inspections shall be performed of the welds for conformance with the applicable drawings for package approval. Nondestructive examination as indicated on the applicable drawings for package approval shall be performed with personnel qualified and certified in accordance.

with SNT-TC-1A.

A pressure test shall be performed on the first cask assembly fabricated. The test shall be performed in accordance with ASME B&PV Code,Section III, Subsection NB, Paragraph NB-6200. Leakage tests shall be performed on the cask containment boundary prior to first use. The fabrication verification leakage test can be separated into four tests: (i) Base metal integrity to evaluate the material of the cask body, primary lid ram access cover plate, and the lid port plug, (ii) lid port plug seal integrity, (iii) primary lid inner seal integrity, and (iv) Ram access cover plate inner seal integrity. This testing shall be performed in accordance with written procedures and conform to the requirements of ANSI N14.5. Personnel performing the tests shall be qualified and certified in leakage testing in accordance with SNT-TC-1A. The acceptance criterion requires each component to be individually leak tight, that is, the leakage rate must be less than 10-7 ref-cm3/sec with a sensitivity of 10-8 ref-cm3/sec, or better.

The base metal of structural and containment components is inspected and tested in accordance with the applicable sections of the ASME B&PV Code. Code alternatives are provided in an appendix to chapter 9.

The DSCs to be transported were all previously fabricated, loaded, and maintained under 10 CFR Part 72 requirements. Verification that previously loaded DSCs and fuel assemblies are compliant with the 10 CFR Part 71 requirements is included in chapter 8.

A thermal test shall be performed to measure the effective thermal conductivity of the assembled cask in the radial direction.

After the cask is placed into service, periodic inspections and tests are required to ensure the cask remains acceptable for usage. Maintenance leakage testing shall be performed on the primary lid inner O-ring seal, ram access cover plate inner O-ring seal, lid port plug metal seal, sealing surfaces associated with the above seals, cask body base metal, primary lid base metal, ram access cover plate base metal, and lid port plug base metal. This testing is to be performed in accordance with written procedures and conform to the requirements of ANSI N14.5.

The staff reviewed the periodic inspection and maintenance requirements for the impact limiters and confirmed that periodic leak testing is required every 5 years. The staff determined that the inspection requirement for the impact limiters was sufficient to detect damage and prevent

81 corrosion related degradation. Therefore, the staff determined that the applicants assessment was acceptable.

The neutron shielding ring leakage test described in section 9.1 shall be performed after any sealing material has been replaced. The lid port plug metal seal will be replaced each shipment as the loaded cask is being closed.

To ensure adequate fatigue strength is maintained, the cask is limited to 800 one-way shipments (empty or loaded).

CONDITIONS In addition to the package description, drawings and contents, the following conditions were included in the CoC:

Condition No. 7: The TN Eagle SC and TN Eagle LC packages shall:

(a)

Be prepared for shipment and operated in accordance with the Operating Procedures in chapter 8 and (b)

Meet the Acceptance Tests and Maintenance Program of chapter 9.

Condition No. 8: Additional operating requirements of the TN Eagle SC and TN Eagle LC package include:

Verification that the fuel assemblies to be transported in any DSC meet the characteristics, maximum average initial enrichment and burnup combinations for both intact and failed fuel assemblies as stated in appendix 1.6 of the application and are loaded per the fuel qualification tables in appendix 8.7 of the application.

The aging management plan and evaluation for each DSC, or set of DSCs, shall be submitted to the NRC prior to shipment.

The package is transported by rail in an exclusive use conveyance only. To ensure adequate fatigue strength is maintained, the packaging is limited to 800 one-way shipments (empty or loaded).

Lifting the package, while attached to the transport frame, is not authorized. The TN Eagle package is handled solely in a horizontal position.

Condition No. 9: Transport by air is not authorized.

Condition No. 11: the CoCs expiration date is October 31, 2028.

CONCLUSION Based on the statements and representations contained in the application, and the conditions listed above, the staff concludes that the Model No. TN-Eagle package has been adequately described and evaluated and that the package meets the requirements of 10 CFR Part 71.

Issued with Certificate of Compliance No. 9382, Revision No. 0.

v t e Letter Sequence Approval
EPID:L-2021-NEW-0000, Tn Americas LLC Application for Approval of the Tn Eagle-STC Transportation Package (Open)
Initiation Request, Request, Request, Request Acceptance Supplement, Supplement Administration Meeting Questions 13 May 2022 10 March 2023 Answers Results Approval Other: ML21050A191, ML21050A192, ML22319A069, ML23275A039, ML23275A041
MONTHYEAR2022-01-10 [Table View]

ML23275A040

Text

SUMMARY

Description:

Description:

Description:

4.6 Evaluation Findings

5.6 Evaluation Findings

6.0 CRITICALITY EVALUATION

6.8 Findings

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