ML19221B657

Enclosure 2: Safety Evaluation Report (Letter to B. Seawright Certificate of Compliance No. 9378, Revision No. 0, for the Model No. HI-STAR 100MB Package)

ML19221B657
Person / Time
Site:	07109378
Issue date:	08/09/2019
From:	John Mckirgan Spent Fuel Licensing Branch
To:	Seawright B Holtec
Saverot P
Shared Package
ML19221B654	List: ML19221B655 ML19221B656 ML19221B657
References
EPID L-2018-NEW-0000
Download: ML19221B657 (68)

Text

SAFETY EVALUATION REPORT Docket No. 71-9378 Model No. HI-STAR 100MB Package Certificate of Compliance No. 9378 Revision No. 0

TABLE OF CONTENTS

SUMMARY

................................................................................................................................. 1 1.0 GENERAL INFORMATION ................................................................................................ 3 2.0 STRUCTURAL AND MATERIALS EVALUATION ............................................................. 5 3.0 THERMAL EVALUATION ................................................................................................ 24 4.0 CONTAINMENT EVALUATION ....................................................................................... 31 5.0 SHIELDING EVALUATION .............................................................................................. 36

6.0 CRITICALITY EVALUATION

........................................................................................... 44 7.0 PACKAGE OPERATIONS ............................................................................................... 61 8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM............................................. 63 CONDITIONS .......................................................................................................................... 65 CONCLUSION ......................................................................................................................... 66 ii

SAFETY EVALUATION REPORT Model No. HI-STAR 100MB Package Certificate of Compliance No. 9378 Revision No. 0

SUMMARY

By application dated February 16, 2018, as supplemented February 13, March 15 and May 29, 2019, Holtec International requested approval of the Model No. HI-STAR 100MB as a Type B(U)F-96 package. Revision No. 2 of the package application, dated May 29, 2019, superseded in its entirety the application dated February 16, 2018.

The Model No. HI-STAR 100MB package, designed for exclusive use transport of either moderate to high burnup PWR fuel (burnup up to 55 GWd/MTU), has two limiting cavity lengths, designated as XL and SL. The packaging body, comprised of a nickel steel shell welded to nickel steel bottom and top flanges, provides the containment boundary, the helium retention boundary, gamma and neutron shielding and heat rejection capability of the package. The outer surface of the inner shell is buttressed with a layered combination of lead, steel and Holtite B neutron shielding material. The top flange has bolted closure lid(s) with machined concentric grooves for elastomeric seals. The packaging body also features collapsible trunnions.

There is only one multi-purpose canister (MPC) model, the MPC-32M, designated for use with the Type XL package. Fuel spacers may be used to minimize the assembly to MPC lid gap.

With the MPC configuration, the HI-STAR 100MB utilizes a single bolted lid. There are two fuel baskets, the F-24 M and F-32 M. The F-24 M basket has flux traps. With the fuel basket configuration, only applicable to the Type SL package, the HI-STAR 100 MB has two bolted lids, with each lid equipped with two concentric seals.

Two identical AL-STAR impact limiters, fabricated of aluminum honeycomb crush material completely enclosed by an all-welded stainless steel skin, are attached to the top and bottom of the packaging with 16 bolts each. The personal barrier, placed over the package lying in a horizontal orientation during transport, is a packaging component when in use.

The HI-STAR 100MB is designed for maximum heat loads of 32 kW (package with the F-32M or F-24M basket) or 29 kW with the MPC-32M loaded in the package.

The packaging body cavity is approximately 165 3/8 inches long (SL configuration) or 191 1/8 inches long (XL configuration) with respective total lengths of the packaging body of 197 inches or 212 inches, respectively. Both versions have an inside diameter of 68 3/4 inches, and an outer diameter of approximately 99 1/4 inches without impact limiters, and 124 inches with the impact limiters installed. The maximum packaging weights of the SL and XL versions are approximately 238,600 pounds and 246, 240 pounds respectively. The package, as configured for transport, i.e., including impact limiters, weighs from 288,000 lbs (SL version) to 300,000 lbs (XL version).

The package was evaluated against the regulatory standards in 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 thermal protection, containment, shielding, and criticality control under normal and accident conditions.

NRC staff reviewed the application using the guidance in "Standard Review Plan for Transportation Packages for Spent Nuclear Fuel," NUREG-1617, March 2000.

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

References Holtec International Report No. HI-2188080 Safety Analysis Report on the HI-STAR 100MB Package, Revision 2, dated May 29, 2019.

1.0 GENERAL INFORMATION 1.1 Packaging The HI-STAR 100MB packaging consists of the following major components: the packaging body, the multi-purpose canister (MPC) or the fuel basket, the impact limiters, and the personal barrier. The packaging has two limiting cavity lengths, designated as XL and SL.

The packaging body, comprised of a nickel steel shell welded to nickel steel bottom and top flanges, provides the containment boundary, the helium retention boundary, gamma and neutron shielding and heat rejection capability of the package. The containment system consists of the inner shell, bottom and top flanges, top closure lid(s), closure lid inner O-ring seal, vent and drain port cover and inner seals, and bolts for the closure lids and port covers.

The outer surface of the inner shell is buttressed with a layered combination of lead, steel and Holtite B neutron shielding material. The top flange has bolted closure lid(s) with machined concentric grooves for elastomeric seals. The packaging body also features collapsible trunnions.

The MPC, a welded cylindrical structure with flat ends, consists of a honeycombed fuel basket made from panels of Metamic-HT, a baseplate, canister shell, lid and closure ring. Fuel spacers may be used to minimize the assembly to MPC lid gap. There is only one MPC model, the MPC-32M, designated for use with this packaging. With the MPC configuration, applicable only to the Type XL package, the HI-STAR 100MB utilizes a single bolted lid.

The fuel basket, made of Metamic-HT both as a structural and neutron absorber material, exists in two configurations, the F-24 M and F-32 M. The F-24 M basket has flux traps. With the fuel basket configuration, only applicable to the Type SL package, the HI-STAR 100 MB has two bolted lids, with each lid equipped with two concentric seals.

Two identical AL-STAR impact limiters, fabricated of aluminum honeycomb crush material completely enclosed by an all-welded stainless steel skin, are attached to the top and bottom of the packaging with 16 bolts each.

The personal barrier, placed over the package lying in a horizontal orientation during transport, is a packaging component when in use, providing a physical barrier to prevent access to hot areas of the package.

The HI-STAR 100MB is designed for maximum heat loads of 32 kW (package with the F-32M or F-24M basket) or 29 kW with the MPC-32M loaded in the package.

The packaging body cavity is approximately 165 3/8 inches long (SL configuration) or 191 1/8 inches long (XL configuration) with respective total lengths of the packaging body of 197 inches or 212 inches, respectively. Both versions have an inside diameter of 68 3/4 inches, and an outer diameter of approximately 99 1/4 inches without impact limiters, and 124 inches with the impact limiters installed.

The maximum packaging weights of the SL and XL versions are approximately 238,600 pounds and 246, 240 pounds respectively. The package, as configured for transport, i.e., including impact limiters, weighs from 288,000 lbs (SL version) to 300,000 lbs (XL version).

1.2 Contents Intact, moderate to high burnup, up to 55 GWd/MTU, spent PWR UO2 fuel assemblies are transported into the HI-STAR 100 MB with a maximum of (i) 24 or 32 PWR UO2 fuel assemblies in the F-24M or F-32M basket, and (ii) 32 PWR fuel assemblies in the MPC-32M.

Control rods are authorized for transport within spent PWR fuel assemblies. Damaged fuel assemblies, fuel debris, and irradiated non-fuel hardware are not authorized for transportation.

1.3 Materials The materials used in the Model No. HI-STAR 100MB package have generally been previously reviewed by staff for the Model Nos. HI-STAR 100, HI-STAR 60, HI-STAR 180, HI-STAR 180D and HI-STAR 190 packages. The bill of materials adequately defines all construction materials, grades and mechanical properties.

The staff reviewed the material properties in Section 2.2 of this SER and finds that the material properties the applicant used are consistent with the commonly available material data and are conservative. On this basis, the staff determined that the material properties of the packaging materials and the contents are appropriate and acceptable.

1.4 Criticality Safety Index The CSI for the Model No. HI-STAR 100MB package is zero, as an unlimited number of packages will remain subcritical under the procedures specified in 10 CFR 71.59(a).

1.5 Drawings The packaging shall be constructed and assembled in accordance with the following drawings:

(a) HI-STAR 100MB Cask Drawing No. 11070, Sheets 1-7, Rev. 2 (b) F-24M Fuel Basket Drawing No. 11083, Sheet 1, Rev. 1 (c) F-32M Fuel Basket Drawing No. 11082, Sheet 1, Rev. 1 (d) MPC-32M Basket Drawing No. 11084, Sheet 1, Rev. 1 (e) MPC Enclosure Vessel Drawing No. 3923, Sheets 1-9, Rev. 38 (f) HI-STAR 100MB Impact Limiter Drawing No. 11101, Sheets 1-4, Rev. 2

1.6 Evaluation Findings

A general description of the Model No. HI-STAR 100MB package is presented in Section 1 of the package application, with special attention to design and operating characteristics and principal safety considerations. Drawings for structures, systems and components important to safety are included in the application.

The package application identifies the Holtec International Quality Assurance Program for the Model No. HI-STAR 100MB package and the applicable codes and standards for the design, fabrication, assembly, testing, operation and maintenance of the package.

The staff concludes that the information presented in this section of the application provides an adequate basis for the evaluation of the Model No. HI-STAR 100MB package against 10 CFR Part 71 requirements for each technical discipline.

2.0 STRUCTURAL AND MATERIALS EVALUATION 2.1 Description of Structural Design 2.1.1 Discussion The Model No. HI-STAR 100MB package has two limiting cavity lengths, designated as XL and SL. The packaging body, comprised of a nickel steel shell welded to nickel steel bottom and top flanges, provides the containment boundary, the helium retention boundary, gamma and neutron shielding and heat rejection capability of the package. The outer surface of the inner shell is buttressed with a layered combination of lead, steel and Holtite B neutron shielding material. The top flange has bolted closure lid(s) with machined concentric grooves for elastomeric seals. The packaging body also features two collapsible trunnions located on the upper cask body, 180 degrees apart, and attached via the trunnion support structure to the neutron shield ribs. Two additional trunnions are attached near the bottom extremity of the package.

The HI-STAR 100MB impact limiters, comprised of a rigid steel cylindrical core, a steel cylindrical skirt that surrounds the crushable aluminum honeycomb block material and ductile steel fasteners, are referred to as AL-STAR, and have been used in all models of the HI-STAR transportation packages.

2.1.2 Codes and Standards Table 2.1.14 of the application lists the applicable codes and standards used for the various components of the HI-STAR 100MB. The applicant used ASME B&PV Code Section III, Division 1, Subsection NB for the structural components of the containment boundary and ASME Code Section II, NUREG-0612 and 10 CFR71.45(a) for the design of the lifting trunnions.

ASME Code Section II was also used for structural support components of the impact limiters.

The staff notes that the codes and standards are adequate and consistent with the intent of 10 CFR 71.45(a).

2.1.3 Design Criteria The design criteria limits for the containment vessel are summarized in Table 2.1.2 of the application. The applicant stated that the free drop events related to Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC) should also meet the stress intensity limits specified under ASME Code,Section III, Level A and Level D, respectively.

The applicant also stated that (i) the overpack closure lid seals must remain functional under all events to ensure leak tightness of the outer containment system and (ii) the containment boundary material must not be susceptible to brittle fracture. The applicant stated that the same stress intensity requirements also apply to the MPC (inner containment boundary). The limits

that govern the design of the containment shell, the closure flange, the containment baseplate, and the closure lids are presented in Tables 2.1.2 - 2.1.4 of the application.

The applicant defined the package as a special lifting device for critical loads. The design criteria of ANSI N14.6 was used for the lifting trunnions and the threaded lifting attachments of the MPC. Additionally, because the applicant also considers the MPC to be a special lifting device, the design criteria of NUREG-0612 were used: the ultimate strength of the MPC lid material, in which the threaded lifting attachments are located, and of the lifting bolts is required to be greater than 10 times the calculated stress in both of those materials.

For the basket, the applicant established a dimensionless panel deformation limit of where is the maximum deflection of the panel and W is the nominal panel width.

Additionally, the applicant stated that creep deformation must remain negligible, brittle fracture must not occur, tearing mode failure must not occur, the B-10 areal density for meeting subcriticality requirements must be assured, the mechanical strength and physical properties under NCT must be maintained, and the physical material properties of the plates must be maintained under neutron and gamma fluence.

Maximum lead slump values were used as the basis for the shielding calculations. To model the lead slump, the applicant assumed that: (i) 5 inches of lead were removed radially from the outside of the bottom forging gamma shield; and (ii) 2 inches of lead were removed in the top and bottom of the gamma shield in the annular space around the containment shell. The applicant stated that if the lead slump remains within these parameters, the package will still maintain its shielding capability.

The impact limiters are designed to absorb the impact energy during a drop event by plastic deformation and stay attached during all postulated impact events.

In Section 2.11, the applicant stated the design criteria for the spent fuel cladding strain is 1.7%

for Zircaloy cladding and 3.4% for M5 cladding under a vertical drop accident condition. The applicant based the Zircaloy strain limit on a study by the Pacific Northwest National Laboratory (Reference 2.11.4 of the application) and the M5 strain limit on test data from the Proceedings of the 2007 International Light Water Reactor Performance Meeting that indicated that high burnup M5 cladding elongation is twice that of Zircaloy.

The staff reviewed the design criteria for the various components of the Model No. HI-STAR 100 MB and determines that they are acceptable because they are consistent with NUREG-1617 and have been previously accepted by the staff (the Zircaloy strain limit was used for the HI-STAR180D, Docket 71-9367).

The staff reviewed all lifting calculations, but only considered the lifting trunnions in the acceptance of this application which requires the yield strength of the trunnion material to be three times greater than the calculated stress in the trunnions. Because all other lifting operations are conducted outside of 10 CFR Part 71, the staff did not consider the performance of other lifting components relevant to this review.

2.1.4 Loading and Load Combinations Five categories of loads are considered for the analysis of the HI-STAR 100MB package: (i) permanent loads, (ii) design loads, (iii) handling loads, (iv) NCT loads, and (v) HAC loads.

Based on the loads considered, the applicant determined two governing load combinations for NCT hot and cold conditions, designated as N1 (including bolt preload, design internal pressure and normal operating temperature) and N2 (including the free drop from a height of 1 foot, bolt preload and maximum normal operating pressure). For HAC conditions, the applicant applied the above HAC loads sequentially as required by 10 CFR 71.73.

The staff reviewed the loads and load combinations and finds that they are acceptable because they are consistent with Regulatory Guide 7.8, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material. Additionally, the staff accepts the load combinations N1 and N2 as the governing load combinations for the HI-STAR 100MB, as being consistent with the governing load combinations for the HI-STAR 180D which is similar to the HI-STAR 100MB.

2.1.5 Weights and Centers of Gravity Table 2.1.11 of the application lists the weights of the various components of the HI-STAR 100MB, and Table 2.1.13 lists the location of the centers of gravity for the SL and XL versions of the package.

2.1.6 Analytical Approach The applicant used a half-symmetry LS-DYNA model to represent the HI-STAR 100MB package, and modeled the containment using a combination of elements, including 8-node solid elements. The number of layers of the elements are sufficient to adequately capture primary membrane and bending stresses, as well as secondary stresses at locations of structural discontinuity. Where shell elements were used in other components of the HI-STAR 100 MB, the applicant chose 10 integration points through the thickness of the element, which is the maximum number possible for LS-DYNA, and is sufficient to ensure convergence of the solution. Each closure lid bolt is modeled with solid elements and the seals are modeled with linear-elastic solid elements to capture seal unloading. The applicant utilized nonlinear elastic-plastic true stress-strain relationships for the key structural member materials.

To qualify the containment boundary for NCT, the applicant developed a static axi-symmetric finite element model using ANSYS. The applicant used layered Plane42 elements to model the through-thickness behavior of the containment shell and baseplate. In previous transportation packages, the applicant used a two-phase approach for the analysis of the dynamic drop events that involved determining the deceleration force of the package using LS-DYNA, then using that inertial deceleration force as a static load to determine component stresses using ANSYS finite element analysis software. However, for the HI-STAR 100MB, the applicant proposed determining the stresses in critical cask components from LS-DYNA directly in the same manner as for the previously approved HI-STAR 190.

In previous applications, the applicant presented the following benchmarking/validation material as evidence to their ability to obtain accurate simulation results for impact events with LS-DYNA:

• Holtec Report No. HI-2156765, Revision 0, Benchmark LS-DYNA for the Free Drops Involving Steel Casks Without Impact Limiters.

• LS-DYNA was used by the applicant to predict the structural response of the HI-STORM FW dry storage cask, including stresses and strains, for non-mechanistic tip over in the HI-STORM FW FSAR (NRC Docket No. 72-1032).

• Comparison of numerical results from the two-phase approach using LS-DYNA and ANSYS to LS-DYNA directly for the HI-STAR 60, HI-STAR 180 and HI-STAR 180D.

• Adherence to guidelines established by the ASME Section III, Division 1 Special Working Group on Computational Modeling for Explicit Dynamics (Use of Explicit Finite Element Analysis for the Evaluation of Nuclear Transport and Storage Packages in Energy-Limited Impact Events - Draft Guidance Document)

While the staff does not consider the comparison of the results of LS-DYNA to other FEA software to constitute benchmarking per se, it can be considered in the aggregate with other benchmarking activities. In the analysis of previous transportation packages, the applicant used LS-DYNA to characterize the quasi-static response of the package to the impact and input that response into ANSYS for further analysis. For the HI-STAR 100MB, the applicant characterized the quasi-static response of the package to the impact using LS-DYNA but, instead of using ANSYS, the applicant continued the analysis in LS-DYNA.

The staff considers LS-DYNA to be a well benchmarked finite element software package, capable of directly providing stress/strain results for impact analysis, provided a quality model is used to predict package material behavior under an energy-limited impact event. The HI-STAR 100MB follows the same modeling approach of previous HI-STAR models such as the 60, 180, 180D and 190 which were benchmarked, in part, by physical drop test data provided in the original HI-STAR 100 model. The applicant demonstrated the ability to model package behavior, similar to previous models, and validated their ability to physical drop test data.

Because of these considerations, the staff determines that the applicants analytical approach to impact analysis using LS-DYNA is acceptable.

The applicant compares the calculated stresses in the component material with the allowable stress by calculating the Factor of Safety (FS) as shown below:

If the FS is greater than 1.0, this indicates that the calculated stresses are less than the allowable stresses, and the structural performance of the component is adequate for that particular loading case. Conversely, if the FS is less than 1.0, the calculated stresses in the component are greater than allowed by the applicants chosen design criteria. Based on the component, and the value of FS, this may still be acceptable, but the applicant must provide an explanation as to why this is acceptable. If the staff determines that the component can still perform the necessary function with a reasonable level of safety, then the staff may accept the deviation.

The staff has reviewed the package structural design description and concludes that the contents of the application meet the requirements of 10 CFR 71.31.

2.2 Materials Properties and Specifications 2.2.1 Drawings

The applicant provided the drawings for the packaging, the MPC, fuel baskets, and impact limiter in Section 1.3 of the application. The drawings include a parts list that provides the material specification, safety category, and primary function of each component. The drawing details and notes also provide welding, examinations, coating requirements, and specifications for alternative materials. The staff reviewed the drawing content with respect to the guidance in NUREG/CR-5502, Engineering Drawings for 10 CFR Part 71 Package Approvals, and confirmed that the drawings provide an adequate description of the materials and fabrication requirements, and, therefore, the staff finds them to be acceptable.

2.2.1 Code and Standards The material codes and standards are described in Table 2.1.14 and the drawings are included in Section 1.3 of the application. The staff reviewed the materials codes and standards to verify that they are consistent with design codes, when applicable, or are otherwise appropriate for the component.

The packaging containment boundary and the MPC enclosure vessel are constructed in accordance with the ASME Boiler and Pressure Vessel Code (ASME Code)Section III, Subsection NB. The principal cask containment boundary components (shell, flanges, and lid) are constructed of ASME SA-350 Grade LF3 or SA-203 Grade E alloy steels. The lid closure bolts are constructed of nickel alloy grade 718 (ASME SB-637) or age-hardened stainless steel type 630 (ASME SA-564 or SA-705). All of the structural materials for the MPC are Alloy X, which may be ASME SA-240 Grade 304, 304LN, 316, or 316LN. Because the packaging and MPC components designed to ASME Subsection NB are constructed of ASME Section II materials, the staff finds the material standards for these components to be acceptable.

Other components of the packaging, fuel baskets, and impact limiters, were not fully designed, fabricated, and tested to a particular code; however, some elements of the ASME Code were used. For example, ASME Code Section III, Subsection NF was used for the fabrication of some non-containment structural welds of the packaging, MPC, and impact limiter. The staff notes that the applicant used ASME Code Section II materials for all of the important-to-safety structural components of the packaging (principally SA-516 Grade 70 carbon steel) and most of the structural components in the impact limiter (principally SA-516 Grade 70 carbon steel).

The fuel baskets are primarily constructed of Metamic-HT material, which is also used in the HI-STAR 190 transportation package (Docket No. 71-9373). The staffs review of the qualification and testing of this material to ensure it meets its structural and criticality-control functions is documented in this SER. The aluminum shims in the fuel basket are constructed of ASTM B221 aluminum alloy. The impact limiter aluminum crush material does not have a material specification, but rather the honeycomb material deformation properties are established via the allowable crush strength ranges specified in the drawing.

The staff finds the material codes and standards for components not associated with the containment boundary to be acceptable because ASME Code Section II materials are specified for most ITS structural components, and the remaining non-Code materials have appropriate qualification and testing requirements to ensure that components can fulfill their intended functions.

2.2.2 Weld Design and Inspection The applicant summarized the welding requirements in Table 2.1.16 and provided detailed welding criteria in Section 8.1.2 and the drawings in Section 1.3. The fabrication and inspection

of the packaging containment boundary and MPC are performed in accordance with ASME Code Section III, Subsection NB. The non-containment boundary welds of the packaging (dose blocker steel) and the impact limiter backbone are fabricated in accordance with ASME Code Section IX and examined in accordance with Subsection V using Subsection NF acceptance criteria. The staff verified that the weld fabrication and examinations described above are consistent with the recommendations in NUREG/CR-3019, Recommended Welding Criteria for Use in the Fabrication of Shipping Containers for Radioactive Materials (NRC, 1984).

As an alternative to ASME welding criteria, the applicant stated that the MPC lid-to-shell weld will be a partial penetration weld and will be examined using a progressive multilayer liquid penetrant (PT) examination during welding. The multi-layer PT must, at a minimum, include the root and final weld layers and one intermediate PT after each approximately 3/8 inch weld depth has been completed. The staff notes that this alternative to the ASME Code is consistent with the NRC guidance in NUREG-1536 (NRC, 2010) for lid-to-shell welds of MPCs used in the storage of spent nuclear fuel, and it is similarly considered appropriate for MPCs used in transportation.

As described in Sections 2.2.1.1.4 and 8.1.5.5 of the application, the Metamic-HT fuel baskets are welded using a friction stir welding process in conformance to ASME Code Section IX. The Metamic-HT Qualification Sourcebook describes the welding qualification program used to demonstrate the performance of this welding method in the manufacture of fuel baskets. The basket drawings require that all basket structural welds be visually examined per the requirements of ASME Code Section III, Division 1, Subsection NG. The staff notes that the use of the Subsection NG requirements for fabrication is consistent with the recommendations in NUREG/CR-3854, Fabrication Criteria for Shipping Containers (NRC, 1984).

In Section 2.2.1.1.4, the applicant stated that all weld filler materials utilized in the welding of ASME Code components will comply with the provisions in the appropriate Code subsection.

The applicant stated that the minimum tensile strength of the weld wire and filler material (where applicable) will be equal to or greater than the tensile strength of the base metal listed in the ASME Code.

The staff verified that the welding processes described above are consistent with ASME Code design criteria or, alternatively, are appropriately qualified and specified in the drawings.

Therefore, the staff finds the welding design and inspections to be acceptable.

2.2.3 Tensile Properties The applicant provided the mechanical properties of the structural materials in Section 2.2.1.

The staff verified that the properties of the carbon and low-alloy steels used in the HI-STAR 100MB package and impact limiter and the austenitic stainless steels used in the MPC-32M are consistent with ASME Code Section II, Part D. The staff also notes that the properties are largely consistent with those used in the previously approved HI-STAR 190 transportation package.

The staff confirmed that the minimum guaranteed values for Metamic-HT basket material listed in Table 2.2.12 are consistent with the tensile properties of the material in the Metamic-HT Qualification Sourcebook. The staff also confirmed that these properties are consistent with those used in the safety analyses for the HI-STAR 180 and HI-STAR 190 transportation packages previously approved. For the aluminum basket shims, the staff verified that the applicant adequately considered potential aluminum strength decreases at the elevated temperatures during normal conditions of transport.

The material deformation properties of the impact limiter aluminum crush material are established via the allowable crush strength ranges specified in the drawing. As stated in Section 8.1.5.3, verification of crush strengths is performed by crush testing of sample blocks of each batch of crush material.

In summary, the staff verified the material tensile properties used in the structural analyses are consistent with ASME Code values or, alternatively, are appropriately qualified and specified in the application and in the licensing drawings. Therefore, the staff finds the tensile properties to be acceptable.

2.2.4 Fracture Resistance of Ferritic Steels The applicant provided the drop weight and impact testing requirements for ferritic components in Table 8.1.5. The applicant included an option to qualify the impact properties for a lowest service temperature (LST) of -40°F, in addition to -20°F as required by 10 CFR 71.71 and 71.73 for normal and accident conditions, respectively.

As described in Section 2.1.3.1 and Table 2.1.17, the applicant proposed to qualify the fracture toughness of ferritic steels by either (1) following the guidance in Regulatory Guides 7.11 and 7.12 to establish the minimum nil ductility transition temperature (TNDT) criteria for Charpy impact and drop weight testing (NRC, 1991a,b) or (2) establishing TNDT with the fracture initiation criteria described in NUREG/CR-3826 for material greater than four inches thick (NRC, 1984).

As stated in Table 8.1.5 of the application, the optional use of the fracture initiation method requires the components to be volumetrically examined to verify that they do not contain flaws that exceed the criteria in the ASME Boiler and Pressure Vessel Code,Section XI, Table IWB-3510-1, Allowable Planar Flaws. If cask operations experience impact loadings that exceed normal conditions of transport, a re-examination is required.

The staff reviewed the proposed fracture toughness testing to ensure that the testing is capable of evaluating the toughness of ferritic steels under impact events at the LST of -20°F. The staff notes that TNDT determined using the fracture initiation criteria from NUREG/CR-3826 is greater than if TNDT is determined following Regulatory Guide 7.12. Thus, the tradeoff for using a higher TNDT for toughness testing is the need to perform volumetric inspections to verify the absence of flaws. The staff also notes that the proposed qualification of impact properties using the fracture initiation approach is consistent with that previously approved by the NRC for the HI-STAR 180, HI-STAR 180D, and HI-STAR 190 packages. The staff finds the applicants approach to qualify the toughness of ferritic steels to be acceptable because it is consistent with the approaches previously approved and the optional use of the fracture initiation criterion to establish test temperatures (rather than the approach recommended in Regulatory Guides 7.12) is supported by additional inspection requirements that are capable of ensuring that flaws are sufficiently small to preclude crack propagation in an impact event.

Although the applicant included an option to qualify the impact properties for an LST of -40°F, this is beyond the requirements of 10 CFR Parts 71.71 and 71.73, which only require that impact performance be demonstrated to -20°F. As a result, the staffs review focused on whether the material toughness qualification activities for an LST of -40°F in Table 8.1.5 are acceptable to demonstrate performance at an LST of -20°F. The staff finds the proposed testing for an LST of -40°F acceptable because it provides an additional margin to demonstrate reasonable assurance of performance at the required LST of -20°F.

For the containment welds, the applicant proposed not to require drop weight testing, but rather to establish TNDT from the drop weight requirements of the containment shell base material and to use that reference temperature to establish the Charpy impact testing of the welds. The staff notes that this approach is consistent with that previously approved for HI-STAR 180, HI-STAR 180D, and HI-STAR 190 packages. The basis for allowing this ASME Code alternative is documented in an NRC clarification letter (NRC, 2014).

For the ASME SA-564 and SA-705 ferritic steel closure lid bolt materials, Table 8.1.5 includes requirements for Charpy impact testing per ASME Code NB-2333. The staff reviewed the bolt impact testing requirements and finds them acceptable because they are consistent with the ASME Code requirements and are capable of ensuring that the bolts will not be subject to brittle facture in a cask drop or tip-over event.

2.2.5 Fracture Resistance of Metamic-HT Basket Material In Section 2.1.3.2, the applicant provided the basis for the adequate fracture resistance of the Metamic-HT basket material. The application references the proprietary Metamic-HT Qualification Sourcebook, which includes the results of Charpy impact testing, minimum guaranteed Charpy impact properties for production lots, and an analysis of the margin of safety against crack propagation in a drop accident. Based on the applicants demonstration that Metamic-HT has an adequate margin of safety against postulated flaws, and the NRCs prior review and approval of Metamic-HT in the HI-STAR 180 and 190 transportation packages, the staff finds the fracture resistance of this material to be acceptable.

2.2.6 Creep of Fuel Basket Materials In Section 2.1.3.2, the applicant referenced creep test data in the proprietary Metamic-HT Qualification Sourcebook as the basis for adequate resistance of the basket material to creep.

The applicant provided creep calculations performed for the HI-STAR 190 transportation package, which showed that minimal creep is expected. Based on a review of the creep test data in the Holtecs qualification program, and the NRCs prior review of the creep calculations in the HI-STAR 190 transportation package, the staff finds the creep resistance of Metamic-HT to be acceptable.

In Section 2.2.1.2.3, the applicant calculated the cumulative creep strain for the aluminum basket shims. The staff reviewed the applicants creep analysis and confirmed that, using conservative parameters, the cumulative creep strain would not be sufficient to affect the performance of the shims in maintaining the structural support of basket. Therefore, the staff finds that the creep resistance of the aluminum shims is acceptable.

2.2.7 Thermal Properties of Materials In Section 3.2, the applicant described the material properties used in the thermal analyses.

The staff reviewed the thermal conductivity, emissivity, density, and specific heat properties of the package materials provided in Tables 3.3.2 through 3.3.9 and verified that the properties are supported by technical sources (e.g., ASME Code Section II) and are consistent with those used in the previously approved HI-STAR 190. Therefore, the staff finds the thermal properties to be acceptable.

2.2.8 Radiation Shielding Materials As summarized in Section 5.1.1, gamma shielding in the HI-STAR 100 MB package is primarily provided by the overpack body steel, body lead, and the steel lid and base plate. Neutron shielding is provided by the Holtite-B neutron absorber material in the overpack body and bottom flange and Holtite-A or Holtite-B in the top impact limiter. Table 5.3.2 describes the material composition and density values used in the shielding analyses.

The staff reviewed the properties of the metallic materials used in the shielding analyses and verified that the chemical compositions and densities are appropriate. Regarding the Holtite-A and Holtite-B polymeric neutron shielding materials, the staff notes that the use of these materials has been previously reviewed and approved in the HI-STAR 180 and HI-STAR 190 packages. The qualification of these materials, including their performance at elevated temperatures, is described in proprietary Holtec Reports. The staff reviewed the qualification data and thermal analyses and verified that the Holtite materials will remain below their allowable temperatures during normal conditions of transport and during vacuum drying operations. The staff notes that, in fire accidents, the applicant does not credit the Holtite materials in the shielding analysis, and thus the materials exposure to fire temperatures does not need to be considered.

Also, as stated in Section 8.1.5.4, each manufactured lot of Holtite will be tested to verify that minimum boron carbide content, hydrogen content, and density are met. Finally, as stated in Section 8.1.5.7, a shielding effectiveness test must be performed on each cask after its first loading, prior to shipment. Therefore, the staff finds the radiation shielding materials to be acceptable.

2.2.9 Criticality Control Materials In Section 8.1.5.5 and Table 8.1.3, the applicant described the material qualification and testing standards used to verify the structural and neutron absorption performance of the Metamic-HT basket material. Tests are performed on a sampling basis to verify the purity and particle sizes of the raw powder, to verify the boron carbide content of the powder mix via a wet chemistry method, as well as to verify the dimensions, mechanical properties, and boron areal density of the finished panels.

The staff notes that the NRC has previously reviewed and approved the use of Metamic-HT in the fuel baskets of the HI-STAR 180 and 190 transportation packages as well as the HI-STORM 100, HI-STORM FW, and HI-STORM UMAX dry storage systems. The testing plan and acceptance criteria in Table 8.1.3 to verify structural and neutron absorption performance is consistent with that in the HI-STAR 190 application. In addition, the staff verified that the material assumptions used in the criticality analysis are generally consistent with the guidance in NUREG-1536 (NRC, 2010) for dry storage systems.

Because the Metamic-HT validation testing is capable of ensuring that the basket material will meet minimum structural and neutron performance requirements, and the fact that the qualification and validation testing program is consistent with that previously approved in other transportation packages, the staff finds the usage and testing of the Metamic-HT to be acceptable.

2.2.10 Corrosion Resistance In Section 2.2.1.3, the applicant described the corrosion behavior of the package materials. The applicant stated that the exterior surfaces of the cask are coated and the interior surfaces of the cask are lined to preclude significant corrosion. Seating surfaces of the elastomeric lid seal are constructed of, or clad with, stainless steel. The applicant also stated that, because the cask is dried and backfilled with helium, there is no credible potential for corrosion on interior surfaces.

The drawings require all non-stainless steel interior containment boundary surfaces and cask external surfaces to be clad with stainless or coated with a surface preservative. Section 2.2.1.2.5 of the application identifies Thermaline 450, or a chemically identical product, as an allowable preservative.

The staff notes that the materials of construction of the HI-STORM 100MB package are largely consistent with those of the previously reviewed and approved HI-STAR 100, HI-STAR 180, and HI-STAR 190 packages. Similarly, the Themaline 450 coating for non-stainless steel surfaces is consistent with the coatings used in those packages. SAR Section 8.2.2.3 states that all external surfaces of the cask and impact limiters shall be visually inspected for damage prior to each fuel loading, including verification that coatings are intact.

Because the materials of construction are either made with corrosion-resistant stainless steels, or are otherwise clad or coated to prevent corrosion, the staff finds that the applicant adequately designed the package to preclude significant corrosion.

2.2.11 Content Reactions In Section 2.2.1.3, the applicant stated that operating procedures for MPC lid welding include appropriate steps to monitor for the potential accumulation of combustible gases beneath the MPC lid and that, once the MPC is dried and backfilled with helium, there is no longer a potential for hydrogen to be generated due to aluminum-water interaction. The staff reviewed the applicants operating procedures in Section 7.1.6 and verified that the applicant will take appropriate steps to monitor for combustible gases and inert the space under the MPC lid during welding.

In addition, the staff reviewed all the cask and MPC materials with respect to their potential to adversely react with the spent fuel pool environment and, as discussed above in the staffs evaluation of corrosion, the staff finds that the applicant adequately designed the package with the use of stainless steel and appropriate coating of non-stainless steel components to preclude significant content reactions.

2.2.12 Radiation Effects In Section 2.2, the applicant stated that radiation levels will not affect the packaging materials.

The applicant provided estimates of the cumulative gamma and neutron radiation fluence over 50 years to show that levels are orders of magnitude below levels that may be expected to cause degradation to the metallic packaging materials.

The staff evaluated the potential cumulative effects of radiation on metallic storage system components and found that material degradation due to radiation exposure is not credible. The staff notes that studies using simulated reactor irradiation conditions have shown that radiation levels in excess of 1019 n/cm2 are required to produce degradation in the mechanical properties of carbon, alloy, and stainless steels (Nikolaev et al., 2002; Odette and Lucas, 2001; Gamble,

2006). This degradation threshold is several orders of magnitude greater than the expected level of fluence on the transportation package, based on the typical neutron flux of 104-106 n/cm2-s (Sindelar et al., 2011). Because both the applicants and the staffs independent analyses determined that radiation levels are significantly below levels that may degrade material, the staff finds the assessment of radiation effects on metallic components to be acceptable.

The applicant also stated that testing on the Metamic-HT neutron absorber material and the Holtite neutron shielding material shows that they will not degrade over the service life of the package. The staff reviewed the proprietary Metamic-HT Qualification Sourcebook and notes that tests on coupons that were irradiated with an exposure equivalent to 56 years of fluence under design basis fuel storage in the HI-STAR 180 cask did not reveal any changes in the physical condition or neutron attenuation properties. In addition, the qualification of the Holtite neutron shielding materials included radiation exposure testing, which revealed no adverse effects on the condition of the material.

In summary, based on the confirmation that radiation is not expected to reach levels capable of degrading the performance of metallic materials and the Metamic-HT neutron absorber materials, the staff finds that the application adequately accounted for radiation affects in the HI-STAR 100MB design.

2.2.13 Seals For the elastomeric containment seals, Table 2.2.11 provides the required seal characteristics, which includes a minimum radiation tolerance threshold and minimum operating temperatures.

The staff reviewed the seal requirements in the application to verify that the closure seals will be capable of performing their containment function. The staff notes that the SAR requires the use of seals that can operate at minimum sustained and short-term temperatures and under radiation exposure.

The staff reviewed the applicants thermal and shielding analyses and confirmed that the required seal characteristics are adequate for the service environment; therefore, the staff finds the applicants seal specification to be acceptable. Further, the cask maintenance schedule in Table 8.2.1 states that seals are replaced if the seals are found to be damaged, they are not free of debris, excessive compression set is present, leakage rates tests are not passed, or if design life limitations may be exceeded.

2.2.14 Spent Fuel Cladding Integrity Classification The spent fuel contents of the MPC-32M canister and the F-32M and F-24M bare fuel baskets are described in Tables 7.7.1 through 7.7.6 and CoC Section 5(b). The HI-STAR 100MB package contains moderate and high burnup undamaged PWR fuel assemblies.

The applicant defined an undamaged fuel assembly as one without known or suspected cladding defects greater than a pinhole leak or hairline crack and which can be handled by normal means.

The staff evaluated the applicants methodology for classifying the fuel and finds it acceptable because it provides a clear description of the characteristics of the fuel that will be allowed to be transported in the package. Also, the characterization of the fuel is consistent with the guidance in ISG 1, Revision 2 (NRC, 2007).

Mechanical Properties The zirconium fuel cladding, guide tubes, and instrument tubes may be constructed of Zircaloy 2, Zircaloy 4, ZIRLO, or M5. The staff notes that the applicants analysis approach for cladding mechanical properties and performance is identical to that used in the previously approved HI-STAR 190 transportation package.

The staff reviewed the cladding mechanical properties used in Section 2.11 of the application and determined that the applicant used mechanical properties for as-irradiated cladding that reasonably bounds all proposed content alloys (Zircaloy-2, Zircaloy-4, ZIRLO, and M5) based on adequate temperature, cladding hydrogen content, cold work, and fast neutron fluence (burnup equivalent) per the allowable contents.

Consistent with the guidance in ISG-11, Revision 3, the staff verified that the applicant considered an effective cladding thickness that is reduced by an oxide layer. The applicant defined a bounding oxide layer for zircaloy and 40µm for M5 cladding.

The staff finds the bounding oxide thickness to be acceptable, as it is consistent with the results of cladding oxidation research and consistent with values used in the previously reviewed and approved in the HI-STAR 190 transportation package.

In Section 2.11, the applicant defined the allowable strain limits for the cladding. The staff reviewed the allowable strain criteria for the cladding and finds them to be acceptable because they are consistent with the results of cladding research and are consistent with the criteria previously reviewed for the HI-STAR 190 transportation package.

Peak Cladding Temperatures Table 3.2.11 defines the maximum allowable cladding temperatures under drying operations are 400°C (752°F) for high burnup fuel (greater than 45,000 MWD/MTU) and 570°C (1058°F) for moderate burnup fuel (less than or equal to 45,000 MWD/MTU). Thermal cycling during drying operations is limited to the guidance in ISG-11, Revision 3 (less than 10 cycles, with cladding temperature variations less than 65°C (117°F)).

In Section 3.2.2.1, the applicant provided the basis for allowing the cladding temperature for the moderate burnup fuel to exceed 400°C (752°F) during short-term drying operations, citing a Pacific Northwest National Laboratory study that evaluated cladding stresses (Lanning and Beyer, 2004).

The staff reviewed the applicants evaluation and the cited study and finds the applicants approach to be acceptable because the applicant demonstrated that hydride reorientation is not expected to be credible during short-term operations for its moderate burnup fuel.

Cover Gas Consistent with the guidance in ISG-22 (NRC, 2006), the staff verified that the applicant will ensure that the fuel rods will be maintained in an inerted environment to prevent oxidation of any potentially exposed fuel pellets.

The operating procedures in SAR Section 7.1.7 requires that the fuel cladding is not to be exposed to air at any time during loading operations.

2.2.15 Bolting Materials The important-to-safety bolts in the HI-STAR 100MB package include the nickel-alloy or age-hardened stainless steel lid closure bolts, alloy steel port cover bolts, and austenitic stainless steel impact limiter bolts.

Section 8.2.2.4 and Table 8.2.1 of the application state that cask closure fasteners and impact limiter fasteners shall be visually inspected for damage prior to their installation and/or transport.

In addition, bolting of the closure lid and port cover plate shall be replaced on the change-out schedule in SAR Table 8.2.1, which is based on an ASME Code Section III fatigue analysis.

The staff finds that the periodic inspection and replacement of bolts are adequate to ensure that age-related degradation will not challenge the lid bolts containment function.

The staff reviewed the thermal expansion properties of the bolting materials in ASME Code Section II, Part D, and noted that the expansion coefficient of the stainless steel lid bolting material is slightly less than the expansion coefficient for the alloy steel lid. Since a difference in expansion coefficients has a potential to overstress the lid bolts, the applicant analyzed bolt stresses under the scenario of differential thermal expansion.

The staff reviewed the calculation and verified that the bolt stresses generated due to differential expansion remain below allowable stress levels. Therefore, the staff finds that the applicant adequately considered the effects of thermal expansion.

References Adkins HE, Jr, BJ Koeppel, and DT Tang, "Spent Nuclear Fuel Structural Response when Subject to an End Impact Accident," In 2004 American Society of Mechanical Engineers, Pressure Vessels & Piping Division, vol. 483, pp. 207-214. 2004. American Society of Mechanical Engineers, New York, NY.

Gamble, R. BWRVIP-100-A: BWR Vessel and Internal Project, Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds. EPRI-1013396.

Palo Alto, California: Electric Power Research Institute. 2006.

Lanning, D. and C. Beyer, Estimated Maximum Cladding Stresses for Bounding PWR Fuel Rods During Short Term Operations for Dry Cask Storage, Pacific Northwest National Laboratory, January 2004.

NRC, NUREG-1536, Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility, Revision 1, 2010.

NRC, NUREG/CR-7189, Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications, 2015.

NRC, NUREG/CR-3854, Fabrication Criteria for Shipping Containers, 1984.

NRC, NUREG/CR-3826, Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Greater than Four Inches Thick, April 1984.

NRC, Interim Staff Guidance-1, Revision 2, Classifying the Condition of Spent Nuclear Fuel for Interim Storage and Transportation based on Function, 2007.

NRC, Interim Staff Guidance-11, Revision 3, Cladding Considerations for the Transportation and Storage of Spent Fuel, 2003.

NRC, Interim Staff Guidance-22, Potential Rod Splitting Due to Exposure to an Oxidizing Atmosphere during Short-Term Loading Operations in LWR or Other Uranium Oxide Based Fuel, 2006.

NRC, Regulatory Guide 7.11, Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask Containment Vessels with a Maximum Wall Thickness of 4 Inches (0.1 m),

June 1991a.

NRC, Regulatory Guide 7.12, Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask Containment Vessels with a Wall Thickness Greater than 4 Inches (0.1 m) but Not Exceeding 12 Inches (0.3 m), June 1991b.

NRC, Letter to S. Anton, Holtec,

Subject:

Request For Clarification Of U.S. Nuclear Regulatory Commission's Position On The Need For Drop Weight Testing For Containment Boundary Weld Material Qualification For The Model No. HI-STAR 180 Package (71-9325) TAC LA0129, December 9, 2014, ADAMS Accession No. ML14330A157.

Nikolaev, Yu., A.V. Nikolaeva, and Ya.I. Shtrombakh. Radiation Embrittlement of Low-Alloy Steels. International Journal of Pressure Vessels and Piping. Vol. 79. pp. 619-636. 2002.

Odette, G.R. and G.E. Lucas. Embrittlement of Nuclear Reactor Pressure Vessels. Journal of Metals. Vol. 53, Issue 7. pp.18-22. 2001.

Sindelar, R.L., A.J. Duncan, M.E. Dupont, P.-S. Lam, M.R. Louthan, Jr., and T.E. Skidmore.

NUREG/CR-7116, Materials Aging Issues and Aging Management for Extended Storage and Transportation of Spent Nuclear Fuel. Washington, DC: U.S. Nuclear Regulatory Commission.

2011.

2.3 Lifting and Tie-Down Standards 2.3.1 Lifting Devices In Reference 2.1.12 of the application, the applicant evaluated all devices or components related to lifting operations including the trunnions, the baseplate and the closure lid lifting holes and bolts. The applicant referred to applicable sections of the HI-STORM 100 FSAR for the evaluation of the MPC lifting evaluation using the threaded holes on the MPC lid. The applicant evaluated the trunnions and the cask lid using the requirements of NUREG-0612 which requires a factor of safety on yield strength of 6 and a factor of safety on ultimate strength of 10.

For the MPC lifting evaluation, the applicant used Regulatory Guide 3.61 for the yielding criteria, which requires a factor of safety of 3 against yielding, and NUREG-0612 for the failure criteria, which requires a factor of safety of 10 against the ultimate strength. These acceptance criteria

exceed those required by 10 CFR 71.45(a) which only requires a factor of safety of 3 against yielding for lifting.

The staff reviewed the pertinent analyses of Reference 2.1.12 of the application, as well as the results provided in Tables 2.5.1 through 2.5.3 as applicable to the yield criteria specified in 10 CFR 71.45(a) for the package only. The calculated stresses in the lifting attachments are within allowable margins; therefore, the staff finds that the package meets the requirements of 10 CFR 71.45(a).

2.3.2 Tie-Down Devices The applicant stated that the package does not incorporate any structural feature that is used as a tie-down device. Additionally, in Section 7.1.5, the applicant stated that the lifting trunnions will be made inaccessible for tie-down by use of a cover or other device that renders them inoperable.

2.4 General Requirements for All Packages 2.4.1 Minimum Package Size The staff finds that the package satisfies the requirements of 10 CFR 71.43(a) for minimum size.

2.4.2 Tamper-Indicating Features The applicant stated, and drawing No. 9848 illustrates, that the upper impact limiter must be removed to gain access to the closure lid bolts. Additionally, the applicant stated that during transport operations, a cover is installed over one of the access tubes for the impact limiter attachment bolts and attached with a wire tamper-indicating seal with a stamped identifier. This seal will indicate whether or not any tampering with the impact limiter has occurred.

The staff reviewed drawing No. 9848 and the transport procedures in Chapter 7 of the application and determined that the package satisfies the requirements of 10 CFR 71.43(b) for a tamper-indicating feature.

2.4.3 Positive Closure The applicant stated that there are no quick disconnect valves in the containment boundary and that the only access to the cask cavity space is through the closure lid which requires a special handling equipment to remove. According to the applicant, the only other opening in the containment boundary is through the vent and drain port which is sealed with a plug and cap, as well as a bolted cover plate as shown in drawing No. 11070. The applicant asserted that, based on the closure system and the analysis for normal and accident condition pressures, the package meets the requirements for positive closure.

The staff reviewed the drawing No.11070 and the applicants analysis for normal and accident pressure conditions and concludes that the containment system is securely closed by a positive fastening device and cannot be opened unintentionally or by a pressure that may arise within the package and therefore satisfies the requirements of 10 CFR 71.43(c) for positive closure.

2.5 Normal Conditions of Transport 2.5.1 Heat The applicant evaluated the HI-STAR 100 MB transportation package for the effects of thermal expansion as a result of Normal (Hot) Conditions of Transport. Based on an initial temperature

of 70°F, and the final temperatures listed in Table 2.6.2, the applicant calculated the thermal expansion of the various components of the HI-STAR 100 MB transportation package and reported the resulting gaps (initial and final) in Table 3.3.6.

The staff reviewed Table 3.3.6 of the application and the thermal expansion calculations, and determines that there are no interference situations among the various components of the HI-STAR 100 MB transportation package due to differential thermal expansion; therefore, there are no induced stresses within the components, and the hot conditions of 10 CFR 71.71(c)(1) do not substantially reduce the effectiveness of the package.

In Section 2.6.1.4.1, the applicant analyzed the containment boundary under load combination N1 using the ANSYS model described in Section 2.1.6 of this SER. In Table 2.6.5, the applicant presents the calculated stresses and associated factors of safety for the components that comprise the containment boundary at a temperature of 450°F. Because the factors of safety for all components are greater than 1.0, the staff determines that the structural performance of the containment boundary is adequate under load condition N1.

2.5.2 Cold The applicant evaluated the package under cold NCT conditions (-40°F) with respect to internal pressure, allowable stresses, bolt stress, and differential thermal expansion. With respect to internal pressure and allowable stresses, the applicant concluded that the internal pressure will decline with decreasing ambient temperature while the material allowable stresses will increase under the same condition. The applicant concluded that decreasing the load and increasing the available strength of the material would result in larger margins of safety than what would be expected for a hot condition.

The staff reviewed the conclusions made by the applicant and determines that the cold conditions of 10 CFR 71.71(c)(2) do not substantially reduce the effectiveness of the package.

2.5.3 Reduced External Pressure The applicant stated that the reduced external pressure equal to 25 kPa (3.5 psia) is bounded by the results of the internal pressure analysis (Load Combination N1). The staff reviewed Load Combination N1 and concludes that it bounds the reduced external pressure condition, and that the reduced external pressure conditions of 10 CFR 71.71(c)(3) do not substantially reduce the effectiveness of the package.

2.5.4 Increased External Pressure The applicant stated that an increase in external pressure of 140 kPa (20 psia) is bounded by the external pressure of 2MPa (260 psia) required by 10 CFR 71.61, and that no additional analysis was required to demonstrate the performance of the package.

Because the requirements of 10 CFR 71.61 bound those of 10 CFR 71(c)(4), the staff concludes that the increased external pressure conditions of 10 CFR 71.71(c)(4) do not substantially reduce the effectiveness of the package.

2.5.5 Vibration and Fatigue 2.5.5.1 Vibration The applicant stated that the lowest frequency of vibration during normal transport conditions occur due to vibrations of the fuel basket cell wall. In reference 2.1.12 of the application, the applicant assumed the cell wall to vibrate as a simply supported beam. In addition, when in horizontal position, the applicant assumed the HI-STAR package as a supported beam at its two

extremities. For both cases, the lowest natural frequency computed were within the rigid range.

The applicant concluded that vibration effects are negligible from a structural integrity point of view. The applicant also determined the natural frequency of the fuel rod by analyzing it as a clamped beam with a length equal to the longest span between adjacent grid spacers. The applicant reported a frequency of 33.9 Hz, and considers the fuel rod to be rigid under NCT and analyzes the cladding under a 5g load. In Table 2.6.9, the applicant reports a safety factor 5.17 against bending.

The staff reviewed the applicants calculations and determines that the large natural frequencies of the basket plates and the cask will preclude any resonance conditions during NCT and that the structural performance of the fuel cladding is adequate under the vibratory loads for NCT.

2.5.5.2 Fatigue The applicant considered cyclic operations of the containment cask using the criteria of ASME B&PV Code Section III, Division 1, Subsection NB-3222.4(d). The applicant evaluated five conditions including Atmosphere to Service Pressure Cycles, Normal Pressure Service Fluctuation, Temperature Difference at Startup and Shutdown, Temperature Difference for Normal Service, and Mechanical Loads to demonstrate that the package was exempt from detailed fatigue analysis. The staff reviewed the applicants evaluation and, because the criteria from NB-3222.4(d) are satisfied, determines that a detailed fatigue analysis is not required on the containment cask.

In Section 2.6.1.3.2, the applicant performed a fatigue analysis on the closure bolts, the closure lid port cover bolts and the containment closure flange internal closure bolt threads.

The table below presents the maximum permissible number of cycles, determined by the applicant, for each of the components.

Maximum Component Permissible Cycles Closure lid Bolts (SA-564 630/705) 225 Closure Lid Bolts (SB-637 N07718) 253 Closure Lid Port Cover Bolts 558 Containment Closure Flange Internal Closure Bolt Threads 1000 The staff reviewed the applicants calculations and determines that the structural performance of the containment boundary is adequate under the fatigue loads for NCT.

2.5.6 Water Spray Because the HI-STAR 100 MB is a large shipping cask, in accordance with RG 7.8, the staff, finds that the water spray test of 10CFR 71.71(c)(6) has no significance in the structural design of the cask and will not substantially reduce the effectiveness of the package.

2.5.7 Free Drop In Section 2.6.1.4 of the application, the applicant analyzed the HI-STAR 100 MB for the 1-foot free drop under hot conditions using LS-DYNA.

The applicant reported a maximum deceleration of 21.15 g and presented the calculated stresses and associated factors of safety for the containment boundary components subjected to a 1-foot drop (load combination N2) in Table 2.6.6 of the application.

Because the factors of safety for all components are greater than 1.0, the staff determines that the structural performance of the containment boundary is adequate for the 1-foot free drop test.

2.5.8 Corner Drop Because the HI-STAR 100 MB is a large shipping cask, in accordance with RG 7.8, the staff, the staff finds that the corner drop test of 10 CFR 71.71(c)(8) is not applicable and will not substantially reduce the effectiveness of the package.

2.5.9 Compression Because the HI-STAR 100 MB weighs more than 11,000 lbs, the staff finds that the compression test of 10 CFR 71.71(c)(9) is not applicable.

2.5.10 Penetration Because the HI-STAR 100MB is a large shipping cask, in accordance with RG 7.8, the staff finds that the penetration test of 10 CFR 71.71(c)(10) has no structural significance and will not substantially reduce the effectiveness of the package.

The staff has reviewed the packaging structural performance under the normal conditions of transport and concludes that there will be no substantial reduction in the effectiveness of the packaging and that it satisfies the requirements of 10 CFR 71.51(a)(1).

2.6 Hypothetical Accident Conditions The applicant evaluated the structural performance of the HI-STAR 100MB under hypothetical accident conditions based on the sequential application of the tests specified in 10 CFR 71.73.

2.6.1 Free Drop For the free drop, the applicant evaluated different drop scenarios including: horizontal drops, oblique drops, bottom end vertical drop, top end vertical drop and center of gravity over corner drops (CGOC). The applicant summarized the maximum deceleration values for each of the simulated drop scenarios as well as the maximum calculated crush depth of the impact limiter in Table 2.7.2.

In all cases the maximum crush depth of the impact limiter was less than the allowable crush depth. The allowable crush depth is based on the distance from the outside edge of the impact limiter to the closest point on the steel cask or the impact limiter backbone. Table 2.7.3 of the application lists the maximum calculated stress in the components that make up the containment portion of the cask along with their respective allowable stress values, the factors of safety and the governing accident that produced the maximum stress in that component. All safety factors in Table 2.7.3 are greater than 1.0 with the lowest value of 1.39 corresponding to the stress intensity associated with primary membrane in the MPC Enclosure vessel.

In addition to the stress intensities, in Table 2.7.3, the applicant reported that the maximum lead slump in the bottom forging gamma shield and the containment shell gamma shield were less than the allowable slump with a safety factor of 1.19. The applicant also stated that the lid seals remain sufficiently compressed after the drop accidents.

In Table 2.11.3, the applicant reported the maximum cladding deceleration for the 30-foot drop as well as the peak principal strains in the Zircaloy and M5 fuel cladding. The peak principal strains for the Zircaloy and the M5 cladding were below the allowable strain values and result in factors of safety of 1.42 and 2.87 respectively.

Based on a review of the above analysis and calculated structural performance, the staff has reasonable assurance to conclude that the free drop test of 10 CFR 71.73(c)(1) will not diminish the performance of the HI-STAR 100MB.

2.6.2 Crush Because the HI-STAR 100MB weighs more than 1,100 lbs, the staff finds that the crush test of 10 CFR 71.73(b)(2) is not applicable.

2.6.3 Puncture For the puncture test, the applicant stated that both the HI-STAR 190 and HI-STAR 100MB packages have identical puncture resistance (i.e. same intermediate shell material and thickness). In addition, the applicant stated that the HI-STAR 190 cask is 40% heavier and concluded that the horizontal drop puncture resistance associated with the HI-STAR 190 bounds the results for the HI-STAR 100MB. Similarly, the applicant stated that the heavier and larger lid diameter of the HI-STAR 190 package bounds the HI-STAR 100 MB package in terms of top end drop puncture resistance.

The staff reviewed the applicants analysis regarding puncture resistance for the HI-STAR 100 MB package and the use of an empirical formula, the Nelms equation, to evaluate cask puncture that was previously used in the HI-STAR 100 application approved by the staff. The applicant acknowledges that the Nelms equation is only applicable to lead backed steel shells and is only used as a conservative analysis, in addition to the bounding analysis presented.

Based on a review of the applicants analysis, the staff has reasonable assurance to conclude that the puncture analysis provided of 10 CFR 71.73(c)(3) will not diminish the structural performance of the HI-STAR 100 MB.

2.6.4 Thermal The HI-STAR 100MB thermal evaluation consisted of ensuring the average temperature across any section of the containment boundary material remains below the maximum permissible temperature, the outer surface exposed to the fire does not slump, internal interferences due to thermal expansion do not develop among the internal components, and the cask closure lid bolts do not unload causing leakage from the containment boundary.

The applicant further analyzed the closure bolts, and the internal and external closure bolt threads for stress under differential thermal expansion in Calculation No. 9 of Holtec Report No.

HI-288083. Overall, the applicant demonstrated that there was sufficient margin in all cases and that the containment boundary of the HI-STAR 100MB remains intact under fire accident conditions.

Based on a review of the applicants analysis, the staff has reasonable assurance that the thermal test of 10 CFR 71.73(c)(4) will not diminish the structural performance of the HI-STAR 100 MB.

2.6.5 Immersion - Fissile Material This requirement is bounded by the deep water immersion requirement of 10 CFR 71.61; therefore, the staff concludes that the HAC test requirement of 10 CFR 71.73(c)(5) is satisfied.

2.6.6 Immersion - All packages This requirement is bounded by the deep water immersion requirement of 10 CFR 71.61; therefore, the staff concludes the HAC requirement test of 10 CFR 71.73(c)(6) is satisfied.

The staff reviewed the packaging structural performance under the hypothetical accident conditions and concludes the packaging has adequate structural integrity to satisfy the subcriticality, containment, shielding, and temperature requirements of 10 CFR Part 71.51(a)(2).

2.7 Special Requirements for Irradiated Nuclear Fuel Shipments 2.7.1 Deep Immersion In Section 2.7.7 of the application, the overpack containment boundary of the HI-STAR 100 MB was analyzed for deep immersion. The applicant stated that the external pressure of 290 psi acts in a direction that increases the pressure on the land (the contact surface between the top flange and lid); therefore, in leakage of water from this accident condition is not a concern. The applicant used ASME Code Case N-284 to evaluate the stability of the containment shell in Holtec Report No. HI-218803, Revision 0, and assumed the outer dose blocker parts do not prevent the 290 psi pressure from acting directly on the outer surface of the containment shell.

The applicant determined that the containment shell does not yield or buckle, as a result of this accident condition.

The staff reviewed the applicants analysis of the containment structure for the deep immersion accident and determines that the containment structure meets the requirements of 10 CFR 71.61 for irradiated nuclear fuel shipments.

2.8 Summary of Damage The applicant has demonstrated that the HI-STAR 100 MB is capable of maintaining its structural integrity to meet the requirements of 10 CFR 71.61 for deep water immersion and 10 CFR 71.51(a)(2) for the sequentially applied hypothetical accident tests of 10 CFR 71.73.

Specifically, the analyses show that the HI-STAR 100MB containment space and the MPC will individually remain inaccessible to the moderator under the immersion event of 10 CFR 71.73, which follow the free-drop, puncture-drop and fire tests. The closure lid will maintain a positive contact load with the top flange which enables the seals to remain functional as effective leak barriers to moderator intrusion to the containment cavity. While there is some plastic deformation due to the puncture bar test, there is no penetration of the containment barrier.

Finally, the average basket deflection in the active fuel region is less than the established limit and no damage occurs to the fuel cladding.

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

3.0 THERMAL EVALUATION The purpose of this evaluation is to verify that the HI-STAR 100MB transportation package:

1) provides adequate protection against the thermal tests specified in 10 CFR Part 71, and

2) meets the thermal performance requirements of 10 CFR Part 71 under NCT and HAC.

The thermal review shall ensure that the peak cladding temperatures (PCTs) and package component temperatures are below the required limits and that the temperature gradients within the fuel basket are minimized to reduce the thermal stresses.

3.1 Description of Thermal Design The HI-STAR 100MB package is designed for either bare baskets, F-32M and F-24M with a heat load limit of 32 kW, or an MPC-32M with a heat load limit of 29 kW. The heat generation in each fuel assembly is non-uniformly distributed over the active fuel length to account for design basis fuel burnup distribution, as discussed in Chapter 5 of the application. The HI-STAR 100MB package is designed to safely dissipate heat under passive conditions (no wind).

The fuel basket is engineered with a honeycomb structure of Metamic-HT plates to provide for transmission of heat from the interior of the basket to its periphery. The fuel basket includes shims to minimize the resistance to lateral transmission of heat. In the bare basket configuration, the entire cavity is at the same helium pressure. Heat dissipation from the basket to the cask occurs by a combination of contact heat transfer, convection, conduction and radiation.

In the MPC, the MPCs enclosure vessel pressurized helium serves as the conductive and convective medium for heat rejection. The space between the cask cavity and the MPC also contains slightly pressurized high purity helium.

The applicant described in Section 3.1.4, Governing Regionalization Configuration, that the fuel loading in the HI-STAR 100MB package is permitted under uniform loading configuration wherein all storage cells are generating heat at maximum permissible rate, or under regionalized loading (Table 7.7.5) which allows hot fuel placement in certain regions of the fuel basket without challenging fuel cladding temperatures. The uniform heat load case bounds all regionalized storage arrangements and, therefore, thermal calculations are performed assuming the design heat load being uniformly distributed amongst the fuel assemblies. The applicant provides the calculated maximum temperatures in Tables 3.1.1 for NCT and Table 3.1.3 for HAC. The applicant also summarizes the calculated maximum pressures in Table 3.1.2 for NCT and Table 3.1.4 for HAC.

The HI-STAR 100MB package thermal design safely dissipates heat under passive conditions and the contents temperatures will remain within their allowable values or criteria for NCT and HAC, as required in 10 CFR Part 71. The staff reviewed the thermal design and determined that description of thermal design is appropriate for a thermal evaluation.

3.2 Material Properties and Component Specifications The applicant provided the packaging materials thermal properties in Tables 3.2.1 through 3.2.9 and listed the temperature limits of materials, fuel cladding, and packaging components in Tables 3.2.10, 3.2.11, and 3.2.12, respectively. The applicant noted in Table 3.2.12 that (i) no reduction in Holtite-B heat conduction effectiveness is assumed during the 30-minute fire and the air conductivity is assigned to Holtite-B during post-fire cooldown to maximize the calculated fuel cladding and component temperatures under HAC, and (ii) the short-term and fire accident limits for the gamma shield, made of lead, are set below the melting point of lead.

The applicant stated in Section 3.2.2, Component Specifications, that the cladding temperature limits, specified in ISG-11 Rev. 3, are adopted for moderate burnup fuel (MBF) and high burnup

fuel (HBF). Neutron absorber material (Metamic-HT) used for criticality control is stable in excess of 538°C (1000°F).

The staff reviewed the material properties of both the contents and packaging components, and Tables 3.2.10 through 3.2.12 for materials and component temperature limits and agrees with the material properties, component specifications and temperature limits used in the thermal analysis.

3.3 General Considerations for Thermal Evaluations 3.3.1 Thermal Model The applicant evaluated the thermal performance of the HI-STAR 100MB package using the 3-D FLUENT CFD code, as described in Section 3.3.1, Overview of the Thermal Model. The thermal evaluations were performed on the HI-STAR 100MB package for both the baskets (F-32M, 32 kW) and the MPC (MPC-32M, 29 kW).

Based on the methodology used in the thermal model, the staff confirmed that the methodology, which was reviewed and accepted by the NRC, is acceptable for the HI-STAR 100MB package thermal analysis.

3.3.2 Screening Evaluations to Determine Limiting Heat Load Pattern The applicant stated in Section S.6.1 of Report HI-2188066 that the steady-state thermal evaluations were performed, as part of screening calculations, for the F-32M, F-24M, and MPC-32M to define the thermally most limiting transport scenario. Based on the calculated results listed in Table S.6.1 for the F-32M and F-24M and Table S.6.2 for the MPC-32M, the applicant concluded that the F-32M results in the maximum PCT and cavity average temperatures, and is thus adopted as the bounding scenario for further evaluations.

The applicant further evaluated two regionalized heat loading scenarios for the F-32M basket and summarized the PCTs in Table S.6.14 at the design basis heat load of 32 kW, and concluded that the regionalized heat load scenarios are bounded by the F-32M at the design basis uniform heat load (32 kW).

The staff reviewed the analysis, compared the peak cladding temperatures in Tables S.6.1, S.6.2, and S.6.14, and finds acceptable the applicants screening evaluations: the regionalized heat load scenarios are bounded by the F-32M at the design basis uniform heat load of 32 kW.

3.3.3 Grid Sensitivity Studies The applicant stated in Section 3.3.2, Description of HI-STAR 100MB 3D Thermal Model, that the HI-STAR100 MB mesh, defined for the thermal analysis, is guided by grid sensitivity studies carried out in the HI-STAR 190 docket. In this manner, a grid independent calculation of the peak cladding temperature is reasonably assured.

The staff compared the HI-STAR 100MB and the HI-STAR 190 applications in their package configurations and thermal design features and finds reasonable the grid sensitivity analysis performed for the HI-STAR 190.

3.4 Thermal Evaluations of Loading Operations 3.4.1 Time-to-Boil Limits The applicant stated in Section 3.3.6.1, Time-to-Boil Limits, that water inside the package cavity is not permitted to boil during fuel loading or unloading operations, in accordance with NUREG-1536. The applicant performed an adiabatic heat up calculation to determine a bounding heat-up rate based on the package heat load and thermal inertia of the loaded package, and then obtained the maximum permissible time duration for the fuel to be submerged in water. The applicant performed calculations to determine the maximum allowable time for completion of wet transfer operations for the bounding F-32M scenario under a design maximum heat load of 32 kW. The applicant tabulated the maximum allowable time limits for completion of wet transfer operations under flooded HI-STAR 100MB conditions for several decay heat loads as shown in Table 3.3.5.

The staff reviewed Section 3.3.6.1 and Table 3.3.5 of the application, and confirmed that (a) the assumptions and the methodology used to derive the time-to-boil limits are appropriate for the derivation of the maximum allowable time for completion of wet transfer operations (the time-to-boil limit) for the HI-STAR 100MB package, based on staffs engineering justification on physical phenomena and thermal characteristics and (b) the allowable time limits for completion of wet transfer operations of the F-32M, as shown in Table 3.3.5, bound those for the F-24M and MPC-32M, respectively, because of the bounding heat loads of 32 kW in the fuel basket configuration.

3.4.2 Moisture Removal Operations Vacuum Drying (VD)

A transient thermal evaluation is performed under specific heat loads in vacuum conditions.

The time required in cycle 1 for the fuel to heat up from an initial temperature of 100°C (212°F) to 370°C (698°F) for high burnup fuel loading or 540°C (1004°F) for all moderate burnup fuel loading is computed. If drying completion criteria is not met, then the cask cavity or MPC must be backfilled with helium to facilitate cooling and ensure that the steady state maximum fuel temperatures remain below the limits of 400°C for HBF and 570°C for MBF, in accordance with ISG-11 Rev. 3. Up to 9 additional cycles of heat-up and cooldown may be performed for fuel temperature variations of less than 65°C in each cycle. If a total of 10 drying cycles fail to meet the drying criteria, then other drying methods (such as forced helium dehydration) must be used unless the package is defueled.

The staff reviewed the operations and criteria of vacuum drying described in Section 3.3.6.2, and determined that the PCTs of 370°C (698°F) for HBF loaded in the MPC-32M and 540°C (1004°F) for MBF loaded in the F-32M, used for determining the cycle times of drying, are bounding and acceptable, based on the temperature margins allowed for VD operation.

Forced Helium Dehydration (FHD)

The applicant stated in Section 3.3.6.2 that the FHD ensures the PCT below the high burnup fuel temperature limit of 400°C (752°F) for all combinations of spent fuel type, burnup, decay heat and cooling times authorized for loading in the HI-STAR 100MB package because the FHD operation induces a forced convection heat transfer and has the PCT lower than the PCT in the quiescent mode of cooling under NCT.

The staff reviewed Section 3.3.6.2 and agrees that FHD can be conducted to maintain the PCT below the 400°C for all fuel types because of its heat removal capability enhanced by forced convection.

3.5 Thermal Evaluation under Normal Conditions of Transport 3.5.1 Heat and Cold The applicant analyzed the thermal performance of the HI-STAR 100MB package with the F-32M basket, as the bounding scenario under NCT, and provided the PCT and maximum component temperatures in Table 3.1.1. The applicant analyzed the thermal performance of HI-STAR 100MB package with the MPC-32M under NCT, and provided the PCT and maximum component temperatures in Table 3.1.6 of the application.

The staff reviewed Tables 3.1.1 and 3.1.6 and confirmed that (a) the heat load pattern of the F-32M is the bounding case for the HI-STAR 100MB package, due to its bounding decay heat, (b) the PCTs of both the F-32M and MPC-32M are below the limit of 400°C per ISG-11, Rev. 3, (c) the maximum temperatures of the containment seals and lid seals for both the F-32M and MPC-32M are below the design limits shown in Tables 3.2.10 and 3.2.12, respectively, and (d) the maximum component temperatures for both the F-32M and MPC-32M are below their design limits, as indicated in Tables 3.2.10, 3.2.11, and 3.2.12, respectively.

Cold Conditions (-40°C or -40°F)

The applicant assumed zero decay heat and no insolation as the bounding conditions for cold evaluation, and stated that the temperature distribution in the HI-STAR 100MB package uniformly approaches the cold ambient temperature of -40°C (-40°F) with all package materials of construction performing their intended function under this cold condition.

The staff reviewed Section 2.6.2, Cold, and Section 3.3.4, plus the fracture toughness test criteria shown in Table 8.1.5, and confirmed that the materials used for the HI-STAR 100MB package are qualified to a lowest service temperature of -40°C (-40°F).

3.5.2 Maximum Normal Operating Pressure (MNOP)

The applicant stated in Section 3.3.5, Maximum Normal Operating Pressure, that the MNOP evaluation is based on the initial maximum backfill pressure, a 3% fuel rod failure in accordance with NUREG-1617, helium radioactive decay, and generation of flammable gases. The applicant calculated MNOPs using the Ideal Gas Law under heat condition of 38°C ambient, still air & insolation, design heat load, and 30% release of the fission gases and 100% release of the rod fill gas from 3% fuel rod failures. The applicant tabulated the fuel cavity MNOPs and maximum gas pressures in the annulus & inter-lid spaces in Table 3.1.2 for the F-32M and MPC-32M.

The staff reviewed the heat conditions used for the pressure calculations, and the NCT MNOPs shown in Table 3.1.2 and agrees that (a) the calculated MNOPs for the F-32M and MPC-32M under a 3% fuel rod failure scenario are bounded by the design basis pressure as shown in Table 2.1.1 and (b) the calculated maximum pressures in the MPC-32M annulus and F-32M inter-lid space are also below the design basis pressures presented in Table 2.1.1. The NCT design basis pressures in Table 2.1.1 were reviewed and accepted by the NRC in a previous application.

3.5.3 Personnel Barrier Evaluation A maximum surface temperature of 145°C (293°F) was computed for the HI-STAR 100MB package under NCT in still air at 38°C and in shade, as shown in Table 3.1.5, and therefore a personnel barrier is required to provide personnel protection and meet the accessible surface temperature limit specified in 10 CFR 71.43(g) for an exclusive use shipment.

The staff reviewed Table 3.1.5 and confirmed that the use of the personal barrier is required with the accessible package surface temperature above the limit of 85°C as defined by 10 CFR 71.43(g) for an exclusive use shipment.

Based on review on Section 3.5, Thermal Evaluation under Normal Conditions of Transport, the staff agrees that the NCT thermal evaluations are in compliance with 10 CFR 71.71.

3.5.4 Fuel Reconfiguration under NCT The fuel assemblies remain intact prior to and during NCT based on the structural evaluations provided in Section 2.11, Fuel Rods, and the defense-in-depth hypothetical fuel reconfigurations considered in the HI-STAR 190 and HI-STAR 180D applications show no significant impact to the containment boundary and fuel assembly due to robust margins in cavity pressures and fuel temperatures.

The staff reviewed Section 3.3.8 and agrees with the applicants statements because (a) the bounding F-32M PCT and maximum component temperatures are below the design limits shown in Table 3.1.1 and (b) the maximum NCT pressures in the fuel cavity (MPC-32M and F-32M), annulus (MPC-32M) and inter-lid space (F-32M) are below the corresponding design limits shown in Table 2.1.1.

3.5.5 Thermal Expansions under NCT The applicant computed thermal expansions for the components of the HI-STAR 100MB package in the radial and axial directions: (a) basket-to-cavity/MPC radial growth, (b) basket-to-cavity/MPC axial growth, (c) fuel-to-cavity/MPC axial growth, (d) MPC-to-package cavity radial growth, and (e) MPC-to-package cavity axial growth, as shown in Section S.6.6 of Report HI-2188066. The applicant provides the NCT thermal expansions of the F-32M and MPC-32M baskets in Table 3.3.6 of the application and Table S.6.9 of Report HI-21888066.

The staff reviewed those Tables and confirmed that the calculated NCT thermal expansions of both the F-32M and MPC-32M are less than the minimum cold gaps because the high conductivity materials (Metamic-HT and low alloy steels) are used to minimize temperature gradients and the package has adequate clearances to allow unrestrained thermal expansions of the package internals under NCT.

3.6 Thermal Evaluation under Hypothetical Accident Conditions 3.6.1 Maximum Temperatures under HAC The applicant described the design basic fire event in Section 3.4.1, Design Basis Fire Event, and the fire conditions in Section 3.4.2: during the 30-minute HAC fire with insolation, transfer of heat from the fire source to the package takes by a combination of radiation and forced

convection heat transfer with (a) a minimum fire emissivity of 0.9 and a lower-bound package absorptivity (0.8), and (b) the forced convection with heat transfer coefficients from the reported Sandia pool fire test data. During the post-fire cooldown, transfer of heat from the package to the ambient combines radiation with the package surface emissivity of 0.66, natural convection between package surface and ambient air, and air conductivity used for conductivity of neutron shielding materials.

The applicant selected the F-32M as the bounding pattern for the HAC thermal analysis because of its bounding heat load, higher NCT fuel cladding and packaging component temperatures. The applicant presented the PCTs and maximum packaging component temperatures in Tables 3.1.3 for the F-32M fuel basket and 3.1.7 for the MPC-32M under the HAC.

The staff reviewed Sections 3.4.1 and 3.4.2, and Tables 3.1.3 and 3.1.7 and confirmed that (a) the HAC initial conditions, fire conditions and thermal model, described in Section 3.4, are consistent with those in previous applications reviewed and accepted by the NRC and therefore are acceptable; (b) the PCTs are below the fuel cladding temperature limits as shown in Table 3.2.11 and in accordance with ISG-11 Rev. 3; (c) the package maximum component temperatures are below the materials temperature limits as shown in Table 3.2.11 and (d) there is no melting of the lead during the HAC fire.

3.6.2 Maximum Pressures under HAC The applicant tabulated the maximum HAC pressures for the F-32M and MPC-32M in Table 3.1.4 which shows (a) the maximum HAC pressures of fuel cavity of the F-32M and MPC-32M under the conditions with no rods rupture and with assumed 100% rods rupture, respectively, and (b) the maximum HAC pressures of the inter-lid space for the F-32M and of the annulus for the MPC-32M.

The staff reviewed the assumptions, boundary conditions, and parameters used in the HAC thermal evaluations and the calculated pressures presented in Table 3.1.4, and confirmed that the maximum package fire accident pressures in the fuel cavity, inter-lid space (F-32M), and annulus (MPC-32M) are below the HAC design basis pressures defined in Table 2.1.1. The HAC design basis pressures in Table 2.1.1 were reviewed and accepted by the NRC in a previous application 3.6.3 Thermal Expansions under HAC The applicant provides the HAC thermal expansions of the F-32M basket and MPC-32M, respectively, in Table S.6.10 and Table S.6.20 of Report HI-21888066.

The staff reviewed those Tables S.6.10 and S.6.20 and confirmed that the calculated HAC thermal expansions of the F-32M and MPC-32M are less than the minimum cold gaps because the high conductivity materials (Metamic-HT and low alloy steels) are used to minimize temperature gradients and the package has adequate clearances to allow unrestrained thermal expansions of the package internals under HAC.

3.7 Evaluation Findings

According to the statements and representations in the application, the staff concludes that the thermal design has been adequately described and evaluated, and that the thermal

performance of the HI-STAR 100MB package meets the thermal requirements of 10 CFR Part 71.

Specific limitations and conditions are specified for approval of this application and the use of HI-STAR 100MB package. These limitations and features are the following:

1) The design heat loads are limited to 32 kW for the F-32M and F-24M baskets and 29 kW for the MPC-32M. The maximum cell heat load complies with Table 7.7.5 regionalized limits.

2) The maximum allowable time limits for completion of wet transfer operations (time to boil limit) for the HI-STAR 100MB are specified in Table 3.3.5.

3) The moisture removal operations of vacuum drying and cyclic vacuum drying should be in accordance with the guidelines of ISG-11, Rev. 3.

4) The drying operations threshold limits to allow the vacuum drying, without the need to prescribe time limits, are 388°C for high burnup fuel (HBF) and 436°C for moderate burnup fuel (MBF) as specified in Tables 3.3.7 and 3.3.8, respectively.

5) The personal barrier is required for the HI-STAR MB100 package with the maximum accessible surface temperature above the limit of 85°C defined by 10 CFR 71.43(g) under an exclusive use shipment.

4.0 CONTAINMENT EVALUATION The objective of the containment review is to verify that the HI-STAR 100MB package design and performance satisfy the containment requirements of 10 CFR Part 71 and will not exceed the allowable radionuclide release rates under NCT and HAC.

4.1 Description of Outer Containment System The HI-STAR 100MB package utilizes a single lid when used with an MPC (MPC-32M) and a double lid when used with a bare basket (F-32M and F-24M). Both configurations consist of an independent redundant double containment to prevent release of radionuclide materials, and both the single lid and each of the double lids are engineered to meet the leak tight criterion of ANSI N14.5 (2014) under NCT and HAC.

The containment boundary of the single lid cask loaded with the MPC (MPC-32M), as shown in Figure 4.1.2, consists of the containment shell, top flange, bottom flange, closure lid port cover plates with port cover bolts and port cover inner seals, outer closure lid with closure lid inner seal and closure lid bolts, and associated containment welds.

The containment boundary of the dual lids cask loaded with the bare basket (F-32M and F-24M), as shown in Figure 4.1.1, consists of the containment shell, top flange, bottom flange, inner closure lid port cover plates (vent and drain ports) with port cover bolts and inner seals, outer closure lid port cover plate (test port) with port cover bolts and inner seal, inner closure lid with closure lid bolts and outer closure lid with closure lid bolts.

The staff reviewed Section 4.1 and the containment boundaries shown in Figure 4.1.2 for either the single lid or dual lid configuration, and confirmed that application provides a complete

description of the containment boundaries, including containment components depicted in Figures 4.1.2 and 4.1.1.

4.1.1 Containment Vessel The containment vessel consists of the containment shell, containment top flange, containment bottom flange, and either a single (outer) closure lid or dual (inner and outer) closure lids.

For the single lid package, the containment vessel houses the MPC (MPC-32M) which contains the spent nuclear fuel assemblies. For the dual lid package, the containment vessel consists of the space created by the inner closure lid and closure lid bolts and the expanded inter-lid space formed with the outer closure lid. The space enclosed by the inner closure lid is used to house the internal basket designs which contain the spent nuclear fuel assemblies.

The staff reviewed Section 4.1.1 and Figures 4.1.1 and 4.1.2 and confirmed that the containment vessel is the primary containment system and that there is no pressure relief device specified on the containment system of the HI-STAR100MB package.

4.1.2 Containment Penetrations The outer containment system penetrations include the closure lid vent port and drain ports for the single lid for the MPC system (MPC-32M), and include both the inner closure lid vent/drain ports and the outer closure lid vent/drain ports for the dual lids for the bare fuel basket casks (F-32M and F-24M). The applicant states, in Section 4.1.2, that each penetration has redundant elastomeric seals and that the containment penetrations are designed and tested to ensure that the radionuclide release rates specified in 10 CFR 71.51(a)(1) and (a)(2) will not be exceeded.

The staff reviewed the discussion of the containment penetrations, provided in Section 4.1.2, and verified that all containment penetrations shown in Figures 4.1.1 and 4.1.2 are adequately described.

4.1.3 Seals and Welds The HI-STAR 100MB package uses a combination of seals and welds to provide containment under NCT and HAC. In addition, the containment remains securely closed and cannot be opened unintentionally or by the internal pressure of the package, as required by 10 CFR 71.43(c).

The applicant also states that: (a) the containment seals are designed and fabricated to meet the design requirements specified in Section 2.1, Structural Design, and (b) the outer containment system welds consist of the welds forming the containment shell, the weld connecting the containment shell to the top flange and the weld connecting the containment shell to the bottom flange, all full penetration welds.

As described in Section 4.1.3, all containment system welds are completed and inspected in accordance with ASME Code Section III, Subsection NB. The staff reviewed the information provided in Section 4.1.3 and finds that the description of the containment seals and welds are consistent with the details shown in the license drawings provided in Section 1.3, License Drawings.

4.1.4 Single Closure Lid - MPC-32M The lid uses two concentric elastomeric seals form the closure with the containment top flange.

The inner seal on the lid demarcates the containment boundary and is tested for leakage through an inter-seal test port which provides access to the volume between the two elastomeric lid seals. The outer elastomeric seal on the outer closure lid provides a redundant closure. The applicant states that closure of the lid vent port and drain port cover plate is done with multiple port cover plate bolts, that the pre-tensioning of the port cover bolts compresses the port cover plate elastomeric seals between the port cover plate and the closure lid to establish containment, and that the torque values, presented in Table 7.1.1, are established to maintain leak-tight containment during NCT and HAC.

The staff verified that the application provides a complete description of the single closure lid -

MPC, and is acceptable.

4.1.5 Dual Closure Lid - F-32M and F-24M 4.1.5.1 Inner Closure Lid - Bare basket The inner closure lid uses two concentric elastomeric seals to form the closure with the containment top flange surface. The inner seal of the inner closure lid is the containment boundary seal and is tested for leakage through an inter-seal test port which provides access to the volume between the two elastomeric lid seals. The outer elastomeric seal on the inner closure lid provides redundant closure. The applicant states that:

(a) closure of the inner lid vent and drain ports is achieved via bolted cover plates with two concentric elastomeric seals. The inner seal in each port cover is the containment boundary seal and is tested for leakage through an inter-seal test port in the port cover which provides access to the volume between two elastomeric port cover seals; (b) the outer elastomeric seals, located under the inner lid vent and drain port covers, provide redundant closure, and (c) the inner lid containment boundary and redundant boundary sealing surfaces are not subject to corrosion due to the presence of the outer lid and inter-lid cavity helium backfill.

The staff verified that the application provides a complete description of the inner closure lid and that it is acceptable for the containment evaluation. The staff also finds that the seal materials will resist corrosion due to the presence of an outer closure lid and the inter-lid cavity helium backfill, which creates an inert atmosphere for the seals.

4.1.5.2 Outer Closure Lid - Bare Basket The outer closure lid uses two concentric seals to form the closure with the containment top flange surface. The inner seal in outer closure lid is the containment boundary seal and is tested for leakage through an inter-seal test port which provides access to the volume between the two elastomeric lid seals. The outer elastomeric seal on the outer closure lid provides redundant closure. The applicant also states that:

(a) closure of the outer lid access port is achieved via a port cover test plug with two concentric elastomeric seals. The port cover test plug inner seal, located at the outer

closure lid port cover plate, is the containment boundary seal and is independently tested for leakage to verify containment performance; (b) a bolted port cover plate, with an outer elastomeric seal, is installed over the port cover test plug to provide redundant closure, and (c) the outer lid containment boundary sealing surfaces are not subject to a corrosive environment due to the presence of redundant closure features that prevent exposure of the seal and sealing surfaces to the environment external to the package.

The staff verified that the application provides a complete description of the outer closure. The staff also finds that the seal materials will resist corrosion due to the presence of redundant closure features.

4.2 Outer Containment System 4.2.1 Normal Conditions of Transport All outer containment system components of the HI-STAR 100MB package are maintained (i) within their code-allowable stress limits and the elastomeric seals remain compressed, and (ii) below their peak temperature and pressure limits. Therefore, the design basis leakage rate will not be exceeded during NCT, as defined in 10 CFR 71.71, because the containment system remains in full compliance with the applicable regulatory temperature and pressure limits.

The applicant states, in Section 4.3.1, that the leak-tight criteria, as specified in ANSI N14.5 (2014), are to be used for all outer containment system leakage tests to ensure that radionuclide release rates specified in 10 CFR 71.51(a)(1) will not be exceeded during NCT.

The staff reviewed Sections 2.6, 3.1 and 4.3, and verified that:

(a) the maximum pressures of 28.9 psig in the annulus space for the MPC-32M and 7.9 psig in the inter-lid space for F-32M are below the NCT design pressure specified in Table 2.1.1, and (b) the peak temperatures of the outer containment system components are below the allowable NCT limits, as shown in Section 3.1.

Therefore, the staff confirmed that the containment integrity of the outer containment system will be maintained under NCT, and the HI-STAR 100MB package meets the containment requirements in 10 CFR 71.71 and 71.51(a)(1).

4.2.2 Hypothetical Accident Conditions All outer containment system components of the HI-STAR 100MB package are maintained (i) within their code-allowable stress limits and the elastomeric seals remain compressed, and (ii) below their peak temperature and pressure limits. Therefore, the design basis leakage rate will not be exceeded during HAC, as defined in 10 CFR 71.73, because the containment system remains in full compliance with the applicable regulatory temperature and pressure limits.

The applicant states, in Section 4.4.1, that the leak-tight criteria, as specified in ANSI N14.5 (2014), shall be used for all outer containment system leakage tests to ensure that radionuclide release rates specified in 10 CFR 71.51(a)(2) will not be exceeded during HAC.

The staff reviewed Sections 2.7, 3.1 and 4.4, and verified that (i) the maximum pressures of 33.4 psig for the MPC-32M in the annulus space and 11.5 psig for the F-32M in the inter-lid space are below the HAC design limits specified in Table 2.1.1, and (ii) all outer containment system component temperatures are below the allowable HAC limits as shown in Section 3.1.

The findings of the structural review confirm that all outer containment system components are maintained within their code-allowable stress limits and that the elastomeric seals remain compressed during HAC. Therefore, the staff confirmed that the integrity of the outer containment system is maintained under HAC and meets the HAC containment requirements in compliance with 10 CFR 71.73 and 71.51(a)(2).

4.3 Inner Containment System 4.3.1 Normal Conditions of Transport All inner containment systems are maintained within their code-allowable stress limits, as shown in Section 2.6 for NCT, and are below their peak temperature and pressure limits, as shown in Section 3.1. Therefore, the design basis leakage rate will not be exceeded during NCT, as defined in 10 CFR 71.71, because the containment system remains in full compliance with the applicable regulatory temperature and pressure limits.

The applicant states, in Section 4.7.2, Containment Criteria, that the leak-tight criterion, as specified in ANSI N14.5 (2014), is applicable for all containment system leakage tests to preclude significant release of radioactive material and ensures that the radionuclide release rates specified in 10 CFR 71.51(a)(1) will not be exceeded during NCT.

The staff reviewed Sections 2.6, 3.1, 4.7.1 and 4.7.2, and verified that (i) the cavity MNOPs of 72.6 psig for the MPC-32M and 55.8 psig for the F-32M, under 3% rod rupture, are below the bounding package cavity MNOP of 100 psig, as specified in Table 2.1.1, and (ii) the peak temperatures of the inner containment system components are below the allowable NCT limits, as shown in Section 3.1.

The findings of the structural review confirm that all outer containment system components are maintained within their code-allowable stress limits and that the elastomeric seals remain compressed during NCT. Therefore, the staff confirmed that the containment integrity of the inner containment system is maintained under NCT and the HI-STAR 100MB package meets the containment requirements of NCT, in compliance with 10 CFR 71.71 and 71.51(a)(1).

4.3.2 Hypothetical Accident Conditions All inner containment system components of the HI-STAR 100MB package are maintained within their code-allowable stress limits and are below their peak temperature and pressure limits. Therefore, the design basis leakage rate will not be exceeded during HAC, as defined in 10 CFR 71.73, because the containment system remains in full compliance with the applicable regulatory temperature and pressure limits.

The applicant states, in Section 4.8.2, Containment Criteria, that the leak-tight criterion, as specified in ANSI N14.5 (2014), is applicable to the inner containment system leakage tests to preclude significant release of radioactive material and ensures that the radionuclide release rates, specified in 10 CFR 71.51(a)(2), will not be exceeded during HAC.

The staff reviewed Sections 2.7, 3.1, 4.8.1, and 4.8.2, and verified that: (i) the maximum cavity pressures of 183.8 psig for the MPC-32M and 183.3 psig for the F-32M, with 100% rods rupture, are below the design limit of 225 psig, as shown in Table 2.1.1, and (ii) all outer containment system component temperatures are below the allowable HAC limits, as shown in SAR Section 3.1.

The findings of the structural review confirm that all outer containment system components are maintained within their code-allowable stress limits and that the elastomeric seals remain compressed during HAC. Therefore, the staff confirmed the integrity of the outer containment system under HAC and that the system meets the containment requirements of HAC, in compliance with 10 CFR 71.73 and 71.51(a)(2).

4.4 Leakage Integrity Tests for Package Overpack The applicant states, in Section 4.5, Leakage Integrity Tests for the HI-STAR 100MB Overpack, that the helium leak test is used to ascertain the integrity of the outer containment boundary in the HI-STAR 100MB package, in accordance with ANSI N14.5 (2014). The applicant provided Table 8.1.2.A and Table 8.1.2.B for leak test requirements of a single lid cask and a dual lid cask. Each table includes information on the test location, components tested, type of leakage test and allowable leakage rate for fabrication, pre-shipment, periodic and maintenance leakage rate tests.

The staff reviewed Section 4.5 and verified that the conditions described for fabrication leakage rate test, pre-shipment leakage rate test, periodic leakage rate test and maintenance leakage rate test are acceptable and consistent with Tables 8.1.2.A and 8.1.2.B.

4.5 Findings

The staff reviewed the containment sections of HI-STAR 100MB package and concludes that:

(1) the package has been described and evaluated to demonstrate that it satisfies the containment requirements of 10 CFR Part 71, and (2) the package is leakage rate tested to the leak-tight criterion when the spent nuclear fuel assembly is loaded.

The HI-STAR 100MB package meets the requirements of 10 CFR 71.51(a)(1) for NCT and 10 CFR 71.51(a)(2) for HAC 5.0 SHIELDING EVALUATION 5.1 Review Objective The objective of this review is to verify that the HI-STAR 100MB package design satisfies the external radiation limit requirements of 10 CFR Part 71 under NCT and HAC. The applicant has based many of the HI-STAR 100MB package components on those of the HI-STAR 190, which has been previously approved by the NRC staff. In this review, staff considered prior review along with design features new to the HI-STAR 100 MB in its evaluation. The staff followed the

guidance provided in NUREG-1617, Standard Review Plan for Transportation Packages for Spent Nuclear Fuel.

5.2 Description of the Shielding Design 5.2.1 Packaging Shielding Design Features Chapter 1 of the application provides a general description of the package design. The applicant included licensing drawings, which are incorporated by reference as part of the CoC, in the application. In the radial direction from the center, the overpack body consists of a stainless-steel lined cavity surrounded by lead gamma shielding, a layer of stainless-steel, neutron shielding material (Holtite A and B), and an external stainless-steel shell. Axially, the bottom consists of stainless-steel with a layer of Holtite A or B sandwiched within one or more stainless-steel clads.

The applicant included engineering drawings and dimensions of all components in Section 1.3 of the application and bill of materials. The applicant provides material properties pertinent to the shielding analysis in Table 5.3.2. The applicant listed the minimal values used to account for tolerances in Table 5.3.4 of the application. The applicant did not account for tolerances for components that the applicant determined had negligible impact on external dose rates.

The staff reviewed Table 5.3.4 and noted that the applicant used either the minimum dimension from the drawings in Section 1.3 of the application or used reduced dimensions (the nominal value less the manufacturing tolerances) when nominal values are shown in the drawings. The applicant did not account for manufacturing tolerances for basket components. Previous staff reviews have confirmed that the fuel basket material has little effect on external dose rates; thus, staff finds that the applicant effectively accounted for all manufacturing tolerances for components that significantly affect external dose rates.

5.2.2 Summary of Maximum Radiation Levels The HI-STAR 100MB is designed for exclusive use shipment on an open conveyance with an enclosure; thus, for the package under NCT the dose rate limits for 10 CFR 71.47(b) apply. For the HI-STAR 100MB the enclosure consists of a personnel barrier that spans between the two impact limiters. The personnel barrier does not encompass the impact limiters. Thus, a limit of 200 mrem/hr applies to the external surface of the impact limiters, the personnel barrier, and the bottom surface of the rail car.

The applicant calculated the dose rates for both a bare basket and a MPC and presented them in Tables 5.1.1-5.1.8. The applicant used the design-basis fuel loadings as determined in Section 7.7 of the application. The applicant modeled the package configurations consistent with the damaged conditions under NCT and HAC as determined by the structural analyses.

The applicants calculation used a mesh tally to determine the dose rate around the package with relatively fine mesh in its shielding analysis models. The fine mesh tally provides a high degree of fidelity that the model can determine the highest total dose rate around the package.

The staff found the applicants shielding model conservatively neglected the separation provided by the personnel barrier and much of the shielding provided by the impact limiters (see SER Section 5.4.1.2). Thus, staff finds reasonable assurance that the dose rates presented by the applicant are reliable and accurate.

5.3 Source Specification The applicant selected the Westinghouse 17x17 (WE 17x17) and B&W 15x15 assemblies as design basis spent fuel assemblies. The applicant chose the B&W 15x15 since it has the

largest fuel mass of the assemblies that may be loaded into the HI-STAR 100MB, and the applicant included the WE 17x17 due to its widespread use and, except for the B&W 15x15, a larger fuel mass than the rest of the assemblies allowed. Staff finds the applicants selection of design basis assemblies acceptable because a larger mass correlates with a larger source term for a given set of burnup, enrichment, cooling time, and irradiation parameters.

The applicant determined both neutron and gamma source terms with TRITON and ORIGAMI modules within the SCALE 6.2.1 code system using the 252-group cross section which is derived from the ENDF/B-VII cross-section library. The applicant assumed a single, full-power cycle to the desired burnup. For ORIGAMI source term calculations, the applicant used predetermined spent fuel assembly characterization database distributed with SCALE 6.2.1.

The applicant does not include source terms for all burnup, enrichment, and cooling time combinations, but rather presents a selected few in Section 7.7 of the application. However, the applicant did analyze all the BECT combinations in their analysis. Given that the source terms are independent of the package design, and prior staff review found the applicants method acceptable, the staff did not review in more details on the applicability of the code.

5.3.1 Gamma Source Gamma radiation in spent fuel originates from decay of actinides and fission products; secondary photons from neutron capture; activation of fuel hardware and non-fuel hardware.

The applicants analysis includes gammas with an energy range from 0.45 to 3.0 MeV in the shielding analyses. The applicant explained that gamma rays above 3.0 MeV are such low intensity and those below 0.45 MeV will not penetrate the steel layers. The staff reviewed the applicants calculated source spectra and finds the applicants choice of the range of gamma energies to be acceptable because it covers the all gammas emitted from the spent fuel and non-fuel hardware that are important to dose rate calculations.

The applicant accounted for activation in the plenum regions. The applicant assumed a maximum Co-59 impurity of 0.5g/kg in the fuel hardware in its shielding analyses. The staff did not consider this assumption acceptable because there is no supporting document to assure that this level of Co-59 impurity is a requirement for the fuel fabricators to obey. However, the staff finds that there are some safety margins to compensate this non-conservative assumption.

In addition, the users will have to assure that the fuel to be loaded meets this Co-60 limit. A pre-ship dose rate measurement will detect if fuel with higher Co-60 level exceeds the assumed level and prevent the users from making the shipment. However, the approval of this assumption does not establish a basis for the future applications. It is the applicants responsibility to provide positive proof the lower Co-59 impurity is used in all of the spent fuel to shipped using this packaging design.

The applicant accounted for the gamma source from the n-gamma reactions of neutrons with the fuel and package materials. The MCNP will account for n-gamma interactions when running in coupled mode (mode n, p). Thus, the staff finds it to be acceptable.

5.3.2 Neutron Source Neutron radiation in spent fuel originates from spontaneous fission; alpha-n reactions; neutrons produced through subcritical multiplication; and gamma-n reactions.

The applicant minimized the assumed enrichment in determining the neutron source.

Spontaneous fission of Cm-244 accounts for approximately 95% of the total neutron source.

Minimizing the enrichment in the model increases the contents of Cm-244. The staff finds this

to be acceptable because it produces a conservative neutron source. Other sources include alpha-n reactions in Cm-244 and neutrons generated from subcritical multiplication reactions.

The applicant presented the results of its neutron source calculations in Table 5.2.6 of the application.

5.4 Model Specification Section 5.3 of the application includes most of the model description as it pertains to the applicants shielding analysis. Additional details are included in the applicants discussion of the shielding methodology in Section 5.4 of the application. The staff reviewed the shielding model, which included comparison of the details on Section 5.3 with drawings in Chapter 1 of the application, and the applicants sample input files. As discussed early in this SER, the applicant appropriately accounted for manufacturing tolerances in its shielding model (see SER Section 5.2.1). The staff finds that the applicants models are consistent with the package performance under NCT and HAC, with exception of the modeling simplifications discussed in as SER Section 5.4.1.2, which the staff finds to be acceptable.

5.4.1 Configuration of Source and Shielding 5.4.1.1 Source Configuration The applicant homogenized the fuel regions within the basket cells and homogenized the end fitting and plenum regions as reduced density steel. Prior staff review has found this acceptable and the staff did not perform further review on this treatment of the fuel in the package.

The applicant addressed reconfiguration of high-burnup (HBU) fuel in Section 1.2.1 of the application and evaluated hypothetical fuel reconfiguration in Section 5.4.5 for the HI-STAR 100MB with an MPC only. The applicant considered the results for design basis conditions, which accounts for clad thinning and hydride formation in HBU fuel per ISG-11, Revision 3, and the applicant used the lower bounds of Zircaloy physical properties to account for uncertainties in HBU cladding in Section 2.11 of the application. The applicant used the same dose adjustments as it used in the HI-STAR 190 shielding analysis.

The staff previously reviewed the applicants handling of reconfiguration in an MPC and found it acceptable. The applicant did not evaluate reconfiguration in bare basket loadings in accordance to the guidance provided in the draft 2015 RIS (ML14175A203) indicating additional reconfiguration analysis may not be necessary provided cladding temperature remains below the maximum value provided in ISG-11, Revision 3.

Staff reviewed the applicants loading procedure and found the maximum parameters in Table 7.1.7 to be consistent with ISG-11, Revision 3. Given that the applicants MPC analysis with clad parameters adjusted for HBU fuel showed no significant clad failure, and the clad temperature in the pool will remain below the maximum limit before bare basket loading, staff finds the applicants reconfiguration analyses to be sufficient.

5.4.1.2 Packaging Configuration The applicant considered neither the personnel barrier nor the transport vehicle in its shielding model. Rather, the applicant considered the outer envelope of the package with impact limiters to be the same as the outer dimensions of the vehicle. Since this assumption ignores the distance provided by the physical presence of the conveyance, the staff finds this simplification to be conservative and acceptable.

For NCT, the applicant credited the neutron absorbing material in the top impact limiter and stainless steel in both the bottom and top impact limiter for 2 m dose rates. The rest of the impact limiter material is neglected (i.e., crush material, impact limiter skirts, and stainless steel impact limiter shell). The staff finds this to be acceptable since the impact limiter is secured in place under NCT and omitting any material from the model will increase the calculated dose rates. The applicant also modeled the height (thickness) of the impact limiters as 1m whereas the actual height is slightly greater. The staff finds this to be another conservative assumption and therefore acceptable since the point is closer to the source than the actual 2 m from the surface of the package.

The applicant determined the dose rates on the surface and 2 m from the surface of the package. Since this selected location will be closer to the source material than the distance from the actual edge of the conveyance required by 10 CFR 71.47, the analyses provide a slightly over-estimate dose rates and thus staff finds this to be conservative and acceptable.

The applicant did not model the bolts used for the closure lid. Since the bolt fills nearly all the space of the hole, and considering the conservative dimensions used in other lid components, the staff finds this approximation will have negligible effect on dose rates.

The applicant omitted the shielding spacer in the model. The staff finds this acceptable as it neglected the shielding material that is present in the package. The applicant omitted the penetrations in and around the lid from its model. The penetrations are not aligned and will be covered with port covers. Under NCT, these penetrations will also be covered by the steel structure of the impact limiter that the applicant conservatively omitted from the model. The staff conducted its own confirmatory analysis and found that the effect on dose rate at 1 m from the closure lid surface is not significant (see Section 5.6 of this SER) even when the package is under HAC at which the impact limiters were assumed to be lost. On this basis, the staff finds that not explicitly modeling the penetrations to be acceptable.

The applicant moved the elevations of the top and bottom trunnions slightly closer to the active fuel region than is provided on the licensing drawings. The staff finds this to be acceptable since the trunnions are a localized area with reduced shielding. Assuming the trunnions to be closer to the active fuel regions will result in an increase in the calculated dose rates because the active fuel regions have much stronger neutron and gamma sources. The applicant did not model any trunnion material that extended beyond the outer diameter of the cask. The staff finds this to be acceptable because this assumption neglects the presence of the materials of trunnions outside the main body of the package.

The applicant modeled an additional 1 cm of material annularly around the outside of the bottom impact limiter lower strongback plate. The staff analysis confirmed this area is of low dose and this material has little impact on the maximum dose rates, although this assumption in not conservative in a general scenario.

The applicant modeled the anti-rotation bar as a part of the surrounding shim than explicitly as a separate component. The staff finds this acceptable since the gaps in question are small internal to the primary shielding components.

The applicant did not model the potential gaps between basket panels. Prior staff review has found the basket has negligible effect on external dose rates. The applicant also did not model the flow-holes in the bottom axial section of the basket. Rather, the applicant reduced the density at that axial location. The staffs confirmatory analysis showed that the dose rate differences are negligible and this approximation to be an acceptable approximation (See SER Section 5.6).

For package under HAC, the applicant assumed three bounding consequences affecting the shielding materials. These effects, which the applicant discusses in Section 2.7 of the application, are: damage to the neutron shield resulting from design basis fire; damage to the impact limiters resulting from a 9 m drop; and lead slump resulting from a 9 m drop. To account for damage to the neutron shield, the applicant removed the neutron shielding material from its HAC model. The applicant also removes all impact limiter material from the accident model.

Since some portion of the neutron material is expected to remain and the impact limiter material was shown to remain in place through calculations in Chapter 2 of the application, the staff finds these assumptions conservatively assume a greater shielding material loss and over-predict dose rates of the package under HAC.

To model the lead slump, the applicant replaces axial sections of both the top and bottom lead shield with void. Since the applicant does not correspondingly increase the density that would occur from such a slump, the staff finds the applicants modeling of lead slump acceptable as it removes material from the model that would otherwise be present and over-predicts dose rates.

Small, localized damage due to pin puncture, and basket deformation, while possible, will negligibly affect the 1 m dose rate. The puncture tests calculation, described in Section 2.7 of the application, demonstrate no complete loss of dose-reduction material is expected. As distance from the localized reduction of shielding is increased, the overall effect is reduced.

Prior staff review (HI-STORM 190) found this assumption to be acceptable. On this basis, the staff did not perform further detailed review.

The applicant identified the radial steel ribs in the neutron absorber as a potential neutron streaming path. Since neutron absorption in steel is less than that of Holtite-B, there is a potential for a localized neutron dose peak. Conversely, gamma attenuation is higher in steel, resulting in a localized gamma dose reduction. The applicant modeled the radial steel ribs within the neutron shield cells. The applicant calculated the dose rates in a fine azimuthal grid near locations aligned with the ribs to capture local dose peaks as a result. The staff reviewed the applicants model and found the azimuthal tally delineations to be sufficiently fine to determine localized effects on external dose rates.

5.4.2 Material Properties The applicant listed the composition and densities of the materials used in the HI-STAR 100MB in Table 5.3.2 of the application. Information on the neutron absorbers, Holtite and Metamic, are given in Chapters 2 and 8 of the application, respectively. The staff reviewed the material properties and the applicants sample input and found them to be appropriately applied in the applicants analysis.

The applicant modeled Holtite-B with minimum thickness, considering tolerances. The applicant also reduced the density of Holtite-B in its model to less than the minimum specified, in part to account for engineered gaps that may occur within the material. In addition, the applicant modeled the B-10 content below that of the minimum requirements. In reducing the B-10 content, the hydrogen fraction of the modeled material in increased, which increases the neutron attenuation in the shielding layer. To compensate this non-conservative assumption, While the hydrogen fraction may be higher, the applicant reduced the density of the Holtite-B such that the overall hydrogen density is still lower than specified in Table 2.2.13 of the application. The staff reviewed the B-10 density and confirmed that, given a minimum B-10 abundance of 19.1%, the applicants B-10 modeling is still below the value resulting from the minimum B4C content specified. The staff finds the applicants adjustment to the material properties of Holtite-B in its shielding analysis acceptable since they all reduce neutron attenuation and absorption and over-estimate neutron dose rates.

The applicant provided a description of the design basis fuel materials the packaging material composition used in its MCNP calculations in Tables 5.3.1 and 5.3.3 of the application, respectively. The application listed the composition of the homogenized fuel region at the end of Table 5.3.2 of the application. The staff reviewed the information and found the composition appropriate. The applicant evaluated both fresh and spent fuel in its MCNP calculations. Fresh fuel is conservative as the subcritical multiplication factor is higher and will result in a larger calculated neutron dose. The applicant modeled spent fuel with the same or bounding burnup and enrichment as the fuel loaded. The applicant also removed neutron absorber materials (e.g., xenon) from its model. The staff finds the material properties used in applicants shielding analyses for the package to be conservative and acceptable.

5.5 Shielding Evaluation The applicant performed its shielding analyses with MCNP-5 Version 1.51. MCNP is a standard Monte Carlo transport code widely used in shielding applications. The applicant determined source terms with the TRITON and ORIGAMI codes from the SCALE 6.2.1 code suite. These codes have been evaluated and proved to be robust methods for determining the radiation sources of spent fuel and activated non-fuel hardware.

5.5.1 Methods The applicant performed separate calculations for each of the radiation sources: neutron; decay gamma; and activation gamma. The applicant based the axial source distribution based on the axial burnup profiles the applicant presented in Table 5.4.4 of the application. The applicant used the same burnup profile from prior applications which staff previously reviewed and found acceptable.

Gamma and neutron source strengths vary with burnup, resulting in higher source strength toward the center of the fuel assembly due to the increased burnup there. In order to account for this effect, the applicant applied scaling factors to 10 axial nodes. The applicant has previously used this method to account for the impact of axial burnup on the distribution of the source term and the staff finds it to be acceptable.

The applicant evaluated the dose rate via a two-step process. The applicants first step is to determine the expected contribution per source particle for each energy group for each of the dose locations. The applicant then scaled the per-particle contribution by the source strength of each energy group at each dose location, added those scaled values to determine a total dose rate at each location, and then normalized the results. The staff finds this acceptable since dose rate scales linearly with source strength for a given shielding design and flux-to-dose rate conversion factors.

The applicant modeled the radial dose locations as a ring around the package. For locations at the flanges where the neutron shield is not present, the applicant used a single 360-degree tally cell. The staff finds this to be acceptable since the calculated value is not likely to vary significantly due to symmetric loading of the package. Since the applicant modeled the steel ribs in the neutron shield, there may be some localized impacts on external dose rates. At these locations, the applicant divided the ring tally cell in 2- and 20.5-degree segments. The smaller azimuthal segments are located outside of the steel ribs in Holtite.

For 1 m and 2 m dose rates, the applicant divided the ring tally into 11.25-degree segments.

The applicant divided the tally axially in 20 cm lengths. The applicant tallied the end dose rates with a circular disk volume that is divided into 20-cm wide radial sections. The staff reviewed the applicants model and found the applicants tallies sufficiently detailed to identify any

differences in dose rates around the package and determine the location at which the highest dose rate would occur.

5.5.2 Uncertainty and Sensitivity The applicant accounts for manufacturing uncertainties by assuming component dimensions are at the most penalizing limit. For shielding evaluations, this is nearly always the minimum value.

Staff finds this acceptable since it minimizes the effectiveness of shielding material during the evaluation.

The applicants source term uncertainty approach is similar to NUREG/CR-6802. The applicant summarized the changes in dose rates compared to the reference cases in Table 5.2.8 of the application. The applicant determined that most source term inputs have little effect on dose rates. However, the applicant found that power density and moderator temperature (density) had a noticeable effect. The applicant accounted for the increased source term due to higher power density by applying a factor, which is a function of cooling time. The staff finds this acceptable as it increases the source term in the applicants model.

The applicant has applied the same conservative PWR burnup profile to account for axial burnup uncertainties in prior applications. Prior staff review found these to be acceptable.

The applicant accounts for MCNP statistical uncertainty by applying one standard deviation to each of the tallies in addition to the conservative assumptions in the underlying model. Staff finds this acceptable since staff has reviewed the applicants MCNP model and finds reasonable assurance that the applicants model is accurate. The applicant presents the list of adjustments in Table 5.4.3 of the application.

5.5.2 Flux-to-Dose-Rate Conversion The applicant used the flux-to-dose rate conversion factors of ANSI/ANS 6.1.1-1977. The staff finds this is consistent with the acceptance criterion provided in NUREG-1617 and therefore to be acceptable.

5.5.4 Radiation Levels The transport index may exceed 10 and the design basis loading is required to be shipped exclusive use.

The applicant shows the calculated dose locations in Figure 5.1.1 of the application. The applicant presented the maximum dose rates determined by the method discussed in SER Section 5.5.1 in Tables 5.1.1 through 5.1.4 of the application. The dose rates presented by the applicant represent the largest value determined over the entire range of the tallies. The staff finds this acceptable because the applicant made conservative assumptions in the calculation and demonstrates that the dose rates meet the regulatory requirements of 10 CFR 71.47 and 71.51.

5.6 Confirmatory Analysis The staff modeled the HI-STORM 100 MB package with all three basket types with the MAVRIC in the SCALE 6.2.1 code suite. MAVRIC is a Monte Carlo transport code. The staff used the ENDF/B-VII multi-group neutron and photon cross-section libraries.

Staff did not homogenize the active fuel region and modeled the fuel as an array of pins in the configuration of the design-basis assemblies. Staff used the applicants design-basis fuel spectrum.

Staff modeled a single rectangular solid to evaluate a single, neutron shield rib location and evaluated any localized effect on external dose rates with point detectors near that location.

Staff evaluated detector locations that the applicant chose and examined a mesh tally to determine if there were any other points of interest. If staffs results showed a location where there might be a higher dose, the staff would generate another importance map and evaluate a new result at that point. This was necessary since variance reduction increases particle tracking near a point of interest at the expense of increased uncertainty at other locations. This apparent dose increase might simply be due to the large statistical error at that other location.

Staff review confirmed the applicants choice of dose locations was appropriate.

Staff made two models to evaluate the presence of flow-holes in the lowest axial level of the fuel basket. One model explicitly models the holes, and another matches the applicants density reduction of that axial level and omits the holes. Staffs analysis did not show significant differences in dose rates between the two models.

Staff also modeled some of the lid penetrations to evaluate the applicants choice not to include them in its model. Staff modeled the penetrations as cylindrical voids of similar dimensions of the vent and lid. Staff analysis found an approximately 10% increase in neutron dose rates at the vent cover by explicitly modeling the lid vents and covers, but at 1 m there was no significant difference in calculated neutron dose rates. The staffs analysis showed a significant difference in gamma dose rate at the surface of the vent cover, however the impact on dose rates becomes less significant as distance increases. Staff analysis found that the presence of the closure lid vent affects gamma dose rate by less than 1.2% at 1 m without the presence of any other shielding material. Under NCT, this location would be covered by the impact limiter material, and the surface dose rate would be measured on the exterior of the impact limiter.

Since 1 m is less than the size of the impact limiter, staff analysis confirms the applicants modeling assumption that omitting the void space resulting from the lid vents does not significantly impact calculated dose rates under NCT. Under HAC, the dose rate is determined from a point 1 m from the surface of the package and given the margin between the applicants calculated dose rates and the regulatory limit in 10 CFR 71.51(a)(2), the staffs analysis confirms that the applicant provided reasonable assurance that the package will meet the regulatory dose limits.

The results of the staffs evaluation confirmed the applicants assumptions and modeling methods are either conservative or do not significantly impact calculated dose rates, considering the margins between the applicants results and the regulatory limits of 10 CFR 71.

5.7 Conclusion Based on its review of the information and representations the applicant provided and the staffs own independent confirmatory calculations, the staff has reasonable assurance that the proposed HI-STAR 100MB package design and contents satisfy the shielding requirements and dose rate limits in 10 CFR Part 71.

6.0 CRITICALITY REVIEW 6.1 Review Objective

The objective of this review is to verify that the HI-STAR 100MB transportation package loaded with the spent fuel assemblies, as specified in the CoC, meets the regulatory requirements of 10 CFR 71.55 and 71.59, i.e., a single package and an array of the packages remains subcritical under NCT and HAC.

6.2 Criticality Safety Evaluation The HI-STAR 100 MB is a Type B(U)F-96 spent nuclear fuel transportation packaging system, designed to transport high burnup PWR UO2 fuel in intact conditions.

The NRC staff reviewed the design of the HI-STAR 100MB package containing the authorized spent fuel. The staffs criticality safety review results are documented in the following subsections of this chapter of this Safety Evaluation Report (SER).

In its review, the staff followed the guidance and the acceptance criteria provided in NUREG-1617, Standard Review Plan for Transportation Packages for Spent Nuclear Fuel (NUREG-1617 thereafter) [Ref. 1].

6.2.1 Package Criticality Safety Design Features The HI-STAR 100MB packaging system consists of an overpack and either a weld-sealed canister (MPC) or an open bare fuel basket installed in inner cavity of the overpack.

The canister and basket designs that can be transported by the HI-STAR 100MB overpack include the MPC-32M, F-32M (bare basket configuration), and F-24M (bare basket configuration). The fuel baskets are constructed with Metamic-HT material, which is an NRC approved boron-doped structural material for fuel baskets. The advantage of using the Metamic fuel basket design is that the fuel is covered by poison plates under both NCT and HAC conditions to assure criticality safety.

The HI-STAR 100MB is designed to transport 32 (MPC-32M and F32-M) or 24 (F-24M) intact PWR fuel assemblies. No damaged fuel is authorized. Table 1.1.2 lists the authorized new MPCs and bare baskets. Chapter 7, Operating Procedures of the application provides specific fuel qualification criteria for the authorized PWR fuel contents.

The fuel basket or canister is a stainless-steel cylinder shell and its internal cavity is compartmentalized into square fuel cells by Metamic-HT plates that contain B-10 as neutron poison. The minimum thickness of the Metamic poison plate is 0.57 inches and the minimum required content of B4C is 10 wt% (weight percent) of the Metamic plate. This value is assured by the acceptance test specified in Table 8.1.3 of the Chapter 8: Acceptance Tests and Maintenance Program.

The HI-STAR 100MB package has two cavity lengths, denoted as Type XL (Extended Length) and Type SL (Standard Length). The Type XL is 191 1/8 inches long and the Type SL is 165 3/8 inches long to accommodate the needs for transporting canisters or fuel assemblies that have the different lengths as specified in Chapter 7 of the application.

Fuel assemblies are held in the fuel cells to maintain their geometric locations in the package under NCT and HAC. Based on structural analyses, all assemblies in the basket are always surrounded by neutron absorbing material to assure criticality safety of the package under NCT and HAC.

The applicant also states that the HI-STAR 100MB package is designed in such a way that the fixed neutron absorber will remain effective for a period greater than 50 years, and there are no credible mechanisms that would cause it to lose its neutron absorption capacity. The staff finds that this is an acceptable determination because depletion of B-10 in the Metamic-HT resulting from neutron irradiation is negligible in comparing to the B-10 load in the material. The staffs

own calculation with a very conservative estimate of the neutron flux from the spent fuel confirmed this conclusion.

The applicant took burnup credit in the criticality safety analyses for the package containing the MPC-32M canister or the F-32M fuel baskets. The applicant states that it used the same burnup credit analysis method as in the previously approved applications for the HI-STAR 190 loaded with the MPC-37 canister and the HI-STAR 100 package as approved in amendment 10 for transport of the MPC-32 canister.

The criticality safety design of the package containing the F-24M basket design includes flux traps but does not take burnup credit. This special design feature reduces the neutronic coupling between the adjacent fuel assemblies in the fuel basket to reduce the potential of nuclear chain reactions so that this basket design does not rely on burnup credit. This is another criticality safety feature that reduces the reactivity, keff, of the package.

The applicant includes the licensing drawings in the application. Licensing drawing No. 11070 shows the layout and dimensions of the HI-STAR 100MB overpack assembly and components.

Licensing drawing 3923 shows the MPC enclosure vessel. Licensing drawing No. 11084 shows the structure and dimensions of the MPC-32M fuel basket. Licensing drawings No. 11082 and 11083 show the structure and dimensions of the F-32M and F-24M fuel basket, respectively.

The HI-STAR 100MB packaging system is designed to transport high burnup fuel (HBU), i.e.,

fuel burnup exceeding 45 GWd/MTU. To support the transport of HBU, the packaging design employs a double containment barrier system design. When the package is loaded with a basket, both the fuel basket lid and the overpack lid are credited for moderator exclusion because both lids are qualified to serve as a stand-alone leak-tight containment boundary.

When the package is loaded with an MPC, the welded fuel basket and the overpack lid closure form two independent containment boundaries for moderator exclusion because both closures are qualified to serve as a stand-alone containment boundary. The structural designs of the basket and overpack ensure that both closures meet the leaktight standard of ANSI N14.5 [Ref.

2] under NCT and HAC as prescribed in 10 CFR 71.71 and 71.73 respectively. Based on the results of structural review, each containment boundary closure meets the water exclusion criterion with significant safety margin because moderator exclusion requires only no ingress of water into the cavity of the package; which is a much less restrictive requirement in comparison with a leak-tight design. On this basis, the staff determined that the package design meets the criterion for moderator exclusion in accordance with the acceptance criteria specified in Interim Staff Guidance number 19 (ISG-19) [Ref. 3].

In summary, the HI-STAR 100MB spent fuel transportation package uses neutron poison plates and burnup credit (MPC-32M and F-32M) or a combination of neutron poison plates and flux trap (F-24M basket only) to ensure criticality safety of the package under NCT. It also employs double closure lids to meet the criterion of moderator exclusion in accordance with the guidance provided in ISG-19 to ensure that the package meets the requirement of 10 CFR 71.55(e), i.e.,

the package remains subcritical under HAC.

6.2.2 Spent Nuclear Fuel Contents The authorized contents include basket designs of the MPC-32M (Metamic HT version of the previously licensed Alloy X-based MPC-32), F-32M (bare basket configuration), and F-24M (bare basket configuration). These baskets are designed to hold 32 (MPC-32M and F32-M) or 24 (F-24M) intact PWR fuel assemblies respectively. No damaged fuel is authorized.

The authorized spent fuel contents include

• Westinghouse (WE) 15x15: 15x15B and 15x15C

• Babcock & Wilcox (B&W) 15x15: 15x15D, 15x15E, 15x15F and 15x15H

• Combustion Engineering (CE) 15x15: 15x15I

• CE 16x16: 16x16A and 16x16B

• WE 16x16: 16x16C

• WE 17x17: 17x17A, 17x17B, 17x17C, 17x17D and 17x17E The design parameters for these authorized spent fuel classes are specified in Tables 7.7.1 and 7.7.2 of the application. Table 7.7.3 (a) provides the minimum burnup for fuel to be loaded in the MPC-32M and F-32M. No minimum burnup is required for fuel to be transported in F-24M basket. Table 7.7.3(b) provides the maximum allowable initial enrichment for each of the fuel classes. Table 7.7.4 provides the bounding fuel depletion parameters.

The non-fuel waste and some non-fuel hardware may be transported in the non-fuel waste basket. However, the total weight of the fissile materials in the wastes is limited to the exempted quantity of fissile materials as defined in 10 CFR 71.15. Therefore, non-fuel hardware in non-fuel waste basket is not subject to criticality safety review. For this reason, the staff did not perform criticality safety review for the packages having non-fuel wastes.

Plutonium fuel in any form other than in the spent UO2 fuel is not authorized for transport.

6.2.3 Summary of Criticality Evaluations The applicant performed criticality safety analyses for the packages containing MPC-32M, F-32M, or F-24M PWR fuel basket respectively. The applicant evaluated a single package with flooded internal and reflected with 30 cm water to demonstrate that the package complies with the regulatory requirements of 71.55(b) and 71.55(d). To demonstrate that the package containing MPC-32M or F-32M basket complies with the regulatory requirement of 71.55(e), that is the package is under HAC, the applicant took credit of the double closure design feature of the package and evaluated the criticality safety without moderator because the package meets moderator exclusion design criterion as provided in ISG-19.

The applicant provides a summary of the calculated keff values in Tables 6.1.1 of the SAR for package containing either the MPC-32M or F-32M basket. The applicant provides a summary of the calculated keff values in 6.1.2 of the SAR for package containing the F-24M basket. The data presented in these tables include the keff values for a single package under NCT and HAC.

The maximum keff value for the package containing the fully loaded MPC-32M or F-32M basket with the maximum U-235 enrichment of 5 wt % is 0.9489. The result for package containing MPC-32M or F-32M basket is based on the credit for a minimum 30.12 GWd/MTU burnup. The data presented in Table 6.1.2 show that the maximum keff value for the package containing the fully loaded F-24M basket with the maximum U-235 enrichment of 4.7 wt% is 0.9496. The calculated keff values show that the package under NCT remains subcritical with all uncertainties in the calculation and an adequate administrative safety margin 0.05k. The staff finds that results demonstrate that the packages containing the MPC-32M, F-32M or F-24M basket meets the acceptance criterion for criticality safety as specified in NUREG-1617. On this basis, the staff finds that the applicant has demonstrated that the package meets the requirements of 10 CFR 71.55(b) and 71.55(d).

The applicant calculated the keff value for a single package containing each subclass of fuel in one of the loading configurations, i.e., MPC-32M, F-32M, and F-24M, as well as an infinite array of packages and provides the results in Table 6.1.3 of the SAR. These data demonstrate that the keff values for the package loaded with different fuel types in the MPC-32M and F-32M are all below 0.95. The staff finds that results meet the acceptance criterion for criticality safety as specified in NUREG-1617.

As discussed earlier in this section of the SER, based on the evaluation results of the packages structural performance under the tests prescribed in 10 CFR 71.73, the staff finds that the applicants criticality safety analysis for the package under HAC is consistent with the damaged condition, i.e., there is no water ingress into the package internal cavity, and the assumptions used in the modeling of the package are conservative and meets the acceptance criterion provided in NUREG-1617. On this basis, the staff finds that the applicant has demonstrated that the package meets the requirements of 10 CFR 71.55(e).

In addition, the applicants criticality safety evaluations show that an infinite array of undamaged or damaged packages remains subcritical. The applicant calculated the criticality safety index (CSI) of this package following the procedures as prescribed in 10 CFR 71.59. The CSI for this package is determined to be 0.

The staff reviewed the applicants calculation of the CSI and modeling of arrays of packages under NCT and HAC. The staff finds that the applicants criticality safety analyses for arrays of packages under NCT and HAC are consistent with the package performance under the respective conditions. And the staff finds that the applicants calculation of the CSI value followed the method prescribed in 10 CFR 71.59 and is therefore acceptable.

6.3 General Considerations for Criticality Evaluations The applicant performed criticality safety analyses for the HI-STAR 100MB package containing either the MPC-32M canister, the F-32M, or package containing the F-24M fuel bare basket loaded with the requested fuel designs. The applicant used the MCNP5.1 [Ref. 4] computer code and the ENDF/B-VII [Ref. 5] cross section library in its criticality safety analyses. For the package taking burnup credit, the applicant used the CASMO5 [Ref. 6] computer code with ENDF/B-VII cross section library to perform depletion analyses to determine the material concentrations of the spent fuel. The staffs evaluations of the computer codes and cross section libraries are documented in Section 6.3.3 of this SER.

The CASMO code is one of widely used computer code for reactor design. The applicant performed code benchmarking analyses for the CASMO5 code using radiochemistry assay (RCA) data from samples of spent fuel from various reactors. Detailed discussion on the use of the CASMO code for burnup credit analysis is discussed in Section 6.6 of this SER.

The applicant applied burnup profile in its burnup credit analysis. The applicant used the axial burnup profiles with 18 burnup axial zones. The staff finds that this detailed modeling of the axial burnup distribution is sufficient to capture the effect of under burned fuel regions at the ends of the fuel assemblies and is consistent with the recommendation of ISG-8, Revision 3

[Ref. 7]. On this basis, the staff finds this approach to appropriate and acceptable.

6.3.1 Model Configuration The applicant explicitly modeled the fuel assembly, Metamic-HT fuel basket fuel cell structure which is a neutron poison material, and other structural and overpack components of the package that are important to criticality safety. The impact limiters are not included in the model.

The package is assumed to contain the most reactive spent fuel authorized to be loaded into each of the specific basket designs, MPC-32M, F-32M or F-24M. The criticality analyses assume 90% of the minimum 10B content in the neutron poison plates manufactured with acceptance criteria as specified in Chapter 8, Acceptance Tests and Maintenance Programs.

The staff finds that this assumption is consistent with the acceptance criterion specified in NUREG-1617 and therefore acceptable.

The fuel stack density is assumed to be 10.686 g/cm3. This is a conservative value because this is 97.5% of the theoretical density of UO2, and it corresponds to a pellet density of 99% or more of the theoretical density. This difference between stack and pellet density is a result of the necessary dishing and chamfering of the pellets. The staff finds that this assumption is conservative and to be acceptable because it is almost the theoretic density, which is the highest possible amount of UO2 fuel per unit volume.

In addition, the applicant used the following assumptions: (1) full flooded package internal with full density of fresh water, i.e., 1.0 g/cc, (2) flooded fuel rod pellet-to-clad gap regions with fresh water under NCT and routine operations (including loading and unloading operations), (3) the worst-case combination of manufacturing and fabrication tolerances, and (4) reflected with 30 centimeters of water outside of the package. These tolerances are consistent with the packages allowable tolerances as shown in the drawings of the package.

The applicant performed sensitivity study on the moderator density variation and finds that the system is most reactive and therefore there is no concern on that preferential flooding may create a more reactive condition. The staffs evaluation of these studies and the conclusion is documented in Section 6.3.4 Demonstration of Maximum Reactivity of this SER.

In the criticality safety analyses for the package, the applicant took credit for 90% of the boron in the Metamic-HT neutron poison plates. The staff finds that this is consistent with the guidance provided in Interim Staff Guidance-23, Application of ASTM Standard Practice C1671-07 when performing technical reviews of spent fuel storage and transportation packaging licensing actions, [Ref. 8] and therefore to be acceptable.

The applicant used fresh fuel composition for the spent fuel assemblies in the package containing the F-24M basket. This assumption does not take credit for the loss of fissile materials and accumulation of fission products and non-fissile transuranic materials that are physically present in the spent fuel assemblies. The staff finds that this is a significant conservative assumption in the criticality safety analysis for the package and therefore to be acceptable.

The applicant performed sensitivity studies on parameters, such as fuel density and water temperature in the cask, that affect reactivity using the CASMO5 code. The results are presented in Table 6.3.4 of the application and show that using the maximum fuel density and the minimum water temperature (corresponding to the maximum water density) provides a bounding condition for criticality safety analyses. The applicant used these conditions in all the criticality safety analysis models. However, the applicant stated that the fuel temperature sensitivity analyses it performed are not used to demonstrate compliance with the regulation; rather they are used to determine sensitivity of the system to changes in temperatures of the moderator, fuel, and structure material and find the bounding conditions. The staff reviewed these analyses and finds it to be acceptable to use the approach for performing comparative analyses because the purpose of this study is to compare the impact of different parameters on the systems reactivity. However, as in indicated by the applicant in the SAR that these analyses are used to identify the bounding parameters rather as the design basis criticality safety analyses which. Computer codes that are used for demonstration of compliance with the regulations need vigorous code benchmarking analyses to identify any potential bias of the code for this specific application.

The applicant took burnup up credit for the packages containing the MPC-32M canister or the F32-M fuel basket that is loaded with the design basis fuel assemblies. The staffs detailed evaluation of the applicants burnup credit analyses is documented in Section 6.3.5 of this SER.

6.3.2 Material Properties The applicant provides material compositions for the various components of the HI-STAR 100MB packaging system in Table 6.3.5 of the application. The data in the table include the nuclide identification number (ZAID) for each nuclide, the atomic number, mass number, and the cross-section evaluation identifier, which are consistent with the ZAIDs in the MCNP manual.

The HI-STAR 100MB package uses Metamic-HT fixed neutron poison plate, which is aluminum alloy containing B4C as neutron absorber. The applicant states that the HI-STAR 100MB is designed to ensure that the fixed neutron absorber will remain effective for a period greater than 50 years and there are no credible degradation mechanisms to cause significant loss of B-10 in the poison plates during this design basis package life-time. The continued effectiveness of the fixed neutron absorber is assured by acceptance testing, documented in Paragraph 8.1.5.5 of the application, to validate the 10B concentration in the fixed neutron absorber. In addition, based on its own calculations for a similar cask model, the staff finds that loss of 10B atoms in the fixed neutron absorber by neutron absorption during the service life time because of irradiation by the neutrons from the content is negligible (less than 10-8 percent of the original loading). Therefore, it is not necessary to provide a surveillance or monitoring program to verify the continued efficacy of the neutron absorber. The applicant provides a detailed physical description, historical applications, unique characteristics, service experience, and manufacturing quality assurance of the fixed neutron absorber to demonstrate that the minimum requirement B-10 concentration is assured. This determination is confirmed by the material reviewer and documented in the material review chapter of this SER. The material compositions and properties of the other packaging materials are consistent with the specifications commonly used criticality safety analyses. However, the Metamic components are still subject to the maintenance requirement for compliance with the regulatory requirements of 10 CFR 71.31(c) and specific maintenance requirements of Chapter 8 of the application.

In the criticality safety analysis models, the applicant did not include the Holtite neutron shield on the outside of the package, nor the impact limiters. The staff finds this modeling simplification to be conservative and acceptable, because these assumptions omit the absorption of neutrons in the neutron shield layer and creates closer reflectors in the radial and axial directions of the package. The combinations of these two factors produces more conservative results in criticality safety analyses. Therefore, the staff did not review the material properties of these components because they are not included in the criticality safety analyses.

The staff reviewed the material properties and the assumptions used in the criticality safety analysis. The staff finds that the material properties the applicant used in the criticality safety analyses are consistent with the commonly available material data and the material properties, such as density, composition, and amount of boron-10. On these bases, the staff determined that the material properties of the packaging materials and the contents are appropriate and acceptable.

6.3.3 Computer Codes and Cross Section Libraries The applicant used the three-dimensional transport theory-based Monte Carlo solution method code MCNP5, Version 5.1 and ENDF/B-VII cross section library in the package criticality safety analyses. In the MCNP criticality safety analysis models, the applicant used 10,000 simulated histories per cycle, a minimum of 400 cycles were skipped before averaging, a minimum of 400

cycles were accumulated, and the initial source was specified as uniform over the fueled regions (assemblies). The applicant explicitly examined the Shannon entropy index, which is part of the MCNP5 model output, to ensure convergence of the calculations by confirming that both the keff value and fission source distribution have properly converged at the end of the calculations. The staff reviewed the Shannon Entropy index provided by the applicant in the sample output file and finds that calculations have adequately converged and therefore the accuracy and reliability of the resultant keff values.

The staff reviewed the applicants criticality safety evaluation method, including the computer code and cross section library as well as the assumptions used in the models. The staff finds that the MCNP code version is one of codes recommended by NUREG-1617 for criticality safety evaluation and the cross-section library represents the up-to-date measurement data. The examination of the Shannon Entropy assured the adequate conversion of the calculations and therefore the accuracy and reliability of the resultant keff values. On this basis, the staff finds that the computer code and cross section library are adequate for this application.

6.3.4 Demonstration of Maximum Reactivity The applicant calculated the neutron multiplication factor, keff, for the HI-STAR 100MB package with each of allowable fuel type as specified in Tables 7.7.1 and 7.7.2 and Chapter 1 of the application. The applicant searched the maximum reactivity with considerations of moderator density and rod pitch changes, which are the two most important parameters that may cause the system keff to change for the package under NCT for some selected enrichment (Holtec Report No: H-2188084). The results are summarized in Table 6.1.3 of the application for single package as well as array of packages under NCT and HAC conditions for each basket type.

Concerning the potential for preferential flooding (also known as partial flooding), the applicant cited the research results published by Cano, et al. [Ref. 9]. The results show that the phenomenon of a peak in reactivity at a hypothetical low moderator density (also called "optimum" moderation) does not occur to any significant extent in a system with heavy neutron poison loading. Based on this publication, the applicant did not evaluate the reactivity of a partially flooded package. Instead, it studied the reactivity effect of variation of the water level in the package. The applicant studied the reactivity changes during the flooding process were evaluated in both the vertical and horizontal positions and provides the results of these calculations are shown in Table 6.3.12. In general, the reactivity increases monotonically as the water level rises, confirming that the most reactive condition is fully flooded. The fully flooded case therefore represents the bounding condition for all basket types. Based on the applicants analyses, as shown in Table 6.3.12 of the application, for package containing either the MPC-32M canister or the F-32M basket, the reactivity increases monotonically as the water level rises, confirming that the most reactive condition is at a fully flooded condition. In these calculations, the cask is partially filled (at various levels) with full density (1.0 g/cm3) water and the remainder of the cask is filled with steam consisting of ordinary water at partial density (0.0002 g/cm3). The staff reviewed the publication cited by the applicant and the staff finds the applicants justification for not performing criticality analyses for preferential flooding of the package to be acceptable because the analyses for the fully flooded package bounds the preferential flooding condition.

The staff reviewed the cited publication and finds that the study is applicable to the HI-STAR 100MB package design because the system is loaded with heavy neutron absorbing materials, namely B-10, in the Metamic plate and is similar to the systems analyzed in the study. On this

basis, the staff finds that there is no need to perform further study on internal preferential flooding of the package.

The applicant studied the reactivity effect of annular fuel pellets (some fuel pellets at the ends of the fuel rods are made with holes in the center of the pellets to save fuel) at the top and the bottom of the fuel rods. Since the annular pellets are loaded at the regions with the high neutron leakage, the applicant determined that the effect on reactivity is minimal. The staff reviewed this conclusion and finds it to be acceptable because these are the regions with the very low neutron importance and their contribution to reactivity is very low and it is consistent with the fundamental of neutron transport and multiplication theory [Ref. 9].

The applicant also studied the reactivity effect of eccentric positioning of assemblies in the fuel cells, i.e., fuel assemblies moving inward toward the center of the fuel basket and presents the results in Table 6.3.13 of the application. The result shows that in most of the cases, moving the assemblies in the regular and specific cells to the periphery of the basket results in a reduction in reactivity, compared to the cell centered position, while moving the assemblies towards the center results in an increase in reactivity, compared to the cell centered position.

All calculations are therefore performed with the assumption that all fuel assemblies moved towards the center of the basket. This result is consistent with the basic nuclear criticality theory that the reactivity increases when the fuel assemblies move toward the center because the neutronic coupling between fuel assemblies gets enhanced. On this basis, the staff finds the results acceptable.

Based on these analyses, the applicant demonstrated that the package under HAC is much less reactive than a fully flooded package even with consideration of highly unlikely fuel reconfiguration. The staff finds the applicants analysis for the package under HAC and the results are consistent with the well understood nuclear physics (much lower reactivity without moderator for low enrichment system) [Ref. 10]. On this basis, the staff determined that the applicant has identified the most reactive configuration of the package and the package meets the regulatory requirements of 10 CFR 71.55(e).

The staff reviewed the method and results of these analyses and finds that the method is acceptable, and the results are reasonable. The staff determined that the applicant has considered all credible NCT and HAC scenarios as prescribed in 10 CFR 71.71 and 71.73 and the method of evaluations and conclusions meet the acceptance criteria of NUREG-1617. On these bases, the staff finds that the package design meets the criticality safety requirements of 10 CFR 71.55(d) and 71.55(e).

6.3.5 Burnup credit analyses for MPC-32M and F-32M packages The applicant took burnup credit for the HI-STAR 100MB package containing either the MPC-32M canister or the F-32M bare fuel basket. The applicant takes credit for all the isotopes as listed in ISG-8, Rev. 3 except Eu-151. The reason is that COSMO code it used to calculate the material composition of spent fuel does not have the capability of tracking this isotope.

The applicant used direct difference method to determine the bias and bias uncertainty of the major actinides as listed in ISG-8, Revision 3. The applicant calculated the bias and bias uncertainty of the major actinides using the direct different method as laid out in NUREG/CR-6811 [Ref. 11]. The applicant used the HTC and MOX critical experiments as recommended by NURG/CR-7109 [Ref. 12] in code benchmarking analysis for the major actinides. The staff reviewed the burnup credit analysis for major actinides and finds that it is consistent with the recommendation with the recommendation of ISG-8, Revision 3 and the results has been

accepted in the previously approved HI-STAR 100 packages. On this basis, the staff finds that the method the applicant used for determining the basis and bias uncertainty for major actinides as listed in ISG-8, Revision 3 to be acceptable.

For the minor actinides and fission products (MAFPs) as listed in ISG-8, Revision 3, the applicant used the recommendation of the ISG and NUREG/CR-7205 [Ref. 13]. In its burnup credit analysis, the applicant used the CASMO-5 code to determine the material composition of the spent fuel with 3 years of cooling time. The applicant then used the MCNP 5.1 computer code to determine the reactivity of the packages.

The applicant states that the results of benchmark evaluations for MPC-37 of HI-STAR 190 are applied to all HI-STAR 100MB fuel packages. The results of burnup credit analysis and verification of assembly burnup evaluated for MPC-37 of HI-STAR 190 are directly applied to the MPC-32M and F-32M fuel packages; however, additional MAFP validation is performed in accordance with ISG-8 Rev. 3 for the MPC-32M and F-32M fuel packages in Appendix 6.C of the SAR.

The applicant also states that for studies that have historically shown a certain aspect or variation is statistically equivalent to the corresponding design basis calculations, or is clearly bounded by those, are not repeated in this chapter. The results of these studies done for MPC-37 of the HI-STAR 190 application are inserted directly in this chapter and their conclusions are applied to the MPC-32M and F-32M baskets. The applicant provides a comparison of the fuel and package design parameters in Table 6.0.2 of the application. It is shown that the differences between the HI-STAR 190 and the HI-STAR 100MB packages are insignificant; therefore, qualitative studies or studies that show insignificant variations are not repeated in this application.

The staff reviewed the application and SER for the HI-STAR 190 and finds the results of the burnup analyses for that package are not applicable to the HI-STAR 100MB because of the significant differences in the capacity, dimensions, and fuel depletion parameters used in the analyses between these two packages. However, the staff did not perform any detailed review on the applicability of the burnup credit analyses for the HI-STAR 190 package because the applicant provided new analyses for the HI-STAR 100MB package.

The applicant developed the fuel loading curves in form of equations as shown in Table 7.7.3(a), which show the minimum required burnup for a given initial enrichment of each PWR fuel class. The loading curves are developed based on a targeted keff value of 0.95. The keff includes all biases and uncertainties at a 95-percent confidence level, should not exceed 0.95.

In the actual calculations, the applicant used a target value of 0.945 as the maximum allowable keff when determining the loading curve for the minimum required burnup as a function of enrichment for MPC-32M canister and F-32M basket. This added another small conservatism in the calculation and safety margin for criticality safety.

The staff reviewed the applicants burnup credit analyses and finds that the method used for the major actinides is acceptable, because it is consistent with guidance and acceptance criteria provided in ISG-8, Revision 3. The applicants use of the recommendation of ISG-8, Rev. 3 for MAFPs, however, is invalid because the depletion analysis computer code CASMO is not capable of tracking all isotopes listed in the ISG and therefore the analyses do not meet the condition of using the recommendation.

Also, the staff finds that the lack of the capability of tracking all isotopes may significantly skew the calculated isotope concentrations of the other isotopes that are included in the burnup analyses because missing one or more major neutron absorber isotope will arbitrarily overestimate some of the isotopes that are tracked and credited.

To assess the overall criticality safety of the package, the staff conducted an independent confirmatory analysis of the reactivity impact of Eu-151 using the SCALE 6.1 package [Ref. 14].

The staff first calculated the Eu-151 concentration based on the design and irradiation parameters of the design basis fuel, including the power density, soluble boron concentration, moderator density, burnup, and cooling time [Ref. 15]. Then the staff built two models, one includes Eu-151 and the other model did not. Based on its own analyses, the staff finds that the estimated Eu-151 concentration is about 0.6 grams per MTU. Comparing the reactivity of the package with and without this isotope, the staff estimated that reactivity is largely comparable with the 1.5% reactivity worth of the MAFPs. On this basis, the staff finds that the burnup credit calculation in this application to be acceptable.

However, it is emphasized that this decision is made based on the staffs own analysis for this specific package design. It does not establish the basis for approval of future applications.

Section 6.6 of this SER provides detailed discussion on the reasons why the recommendation of ISG-8, Revision 3 for the MAFPs is not appropriate when using the CASMO code to calculate the material composition for burnup credit analyses.

6.3.6 Estimated Additional Safety Margin One significant conservative assumption of the burnup credit methodology described here is crediting only recommended set of actinides and fission products, i.e. excluding Eu-151 and other fission products and actinides that are not endorsed by ISG-8. To estimate the safety margin that corresponds to just this single assumption, additional calculations were performed crediting all isotopes from the CASMO5 depletion calculations.

These calculations were performed for Configuration 1, i.e., a uniform loading pattern, and the comparison with the design basis calculations are listed in Table 6.B.6 of the application. The results show that neglecting additional actinides and fission products results in an estimated safety margin of 0.01 delta-k. The staff reviewed the applicants analysis and finds that the result is comparable with similar previously packages with the similar design features. On this basis, the staff determined that the applicants estimation of additional criticality safety margin to be acceptable.

The applicant states that for package with the F-24M basket configuration, there is no need to estimate the additional reactivity margin. The staff finds this assert to be acceptable because this package does not take burnup credit and the guidance of ISG-8 for estimate of additional reactivity is not applicable.

6.4 Single Package Evaluation The applicant performed criticality safety analyses for packages containing the MPC-32M, F-32M, and the F-24M basket under NCT. The results are summarized in Table 6.1.3 for all fuel packages and for the most reactive configurations and fuel condition in each basket. The results include worst combination of manufacturing tolerances, and the computational bias, uncertainties, and computational statistics. The applicant explicitly modeled the fuel basket structure, which also serves as neutron poison plates, and the overpack in the models.

The applicant modeled the fuel assembly assuming flooded fuel cladding gap and the inter-cavity of the package. The applicant ignored the impact limiters. Table 6.1.1 of the application provides the results of the criticality evaluation. The results of the analyses show that the package meets the criticality safety requirements of 10 CFR 71.55(b) and 71.55(d).

The applicant also evaluated the criticality safety of a single package under HAC with the assumptions that the package internal remains dry, i.e., there is no moderator intrusion. The applicant surrounded the package with 30 cm of water reflector. The applicant ignored the impact limiters in the model. Table 6.1.1 of the application provides the results of the criticality calculations. The results show that the package remains subcritical when subjects to the tests prescribed in 10 CFR 71.73.

The staff reviewed the applicants criticality safety analyses for single package under HAC.

Based on the result of structural review as documented in Chapter 2 of this SER, the staff determined that the package meets the acceptance criteria of moderator exclusion design as defined in NUREG-1617 and ISG-19. Therefore, the criticality safety analyses based on moderator exclusion is acceptable. The applicant presents in Table 6.1.1 of the application the maximum keff values for the packages under HAC. The result demonstrates that the package meets the regulatory requirements of 10 CFR 71.55(e). On this basis, the staff determined that the applicant has demonstrated that the package meets the regulatory requirements of 10 CFR 71.55(b), 71.55(d), and 71.55(e).

6.5 Evaluation of Array of Packages under Normal Conditions of Transport and Hypothetical Accident Conditions The applicant performed criticality safety analyses for an array of packages under NCT and HAC separately. The applicant used the same assumptions as it used in the evaluation for single package. Based on its calculations, an array of infinite number of packages under NCT or HAC remain subcritical with considerations of all potential uncertainty and an administrative safety margin of keff = 0.5. Based on the results of the criticality safety analyses for the infinite array of packages under NCT and HAC, the applicant calculated the Criticality Safety Index of the array of packages in accordance to the method prescribed in 10 CFR 71.59 and determined that the CSI for this package is 0.0.

The staff reviewed the applicants analyses of the criticality safety of arrays of packages under NCT and HAC and the calculation of the CSI value. The staff finds that the assumptions used in the models for an array of packages under NCT as well as an array of packages under HAC are conservative and are acceptable based on the acceptance criteria provided in NUREG-1617 and Supplement 1 to NUREG-1617. Based on its review, the staff finds that the package design with the authorized PWR fuel assemblies meets the acceptance criteria as provided in NUREG-1617. The staff also finds that the applicants calculation of the CSI followed the procedures prescribed in 10 CFR 71.59(a) and therefore the calculated CSI value is acceptable.

On these bases, the staff determined that the applicant has demonstrated that the package design meets the regulatory requirement of 71.59.

6.6 Computer Code Benchmarking As discussed earlier in this SER, the applicant used two computer codes CASMO5 and MCNP5.1 in its criticality safety analyses. For the package containing the F-24M bare basket, the applicant assumed that fuel is unirradiated, i.e., fresh fuel assumption. The criticality safety analysis for this package used the MCNP code only because the fuel composition is already known and there is no need for determining the spent fuel composition.

The applicant provides code benchmarking using criticality experiments selected from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments [Ref. 18]. The staff reviewed the list of the critical experiments selected by the applicant and finds these

selected critical experiments are appropriate for the enrichment range and fuel assemblies to be transport. On this basis, the staff determined that the applicants code benchmarking analyses for the package containing spent fuel in the F-24M bare fuel basket is adequate and acceptable.

As discussed earlier in the SER, the package design takes burnup credit for the spent fuel contents loaded in the MPC-32M canister or the F-32M fuel basket. In the criticality safety analysis for these packages, the applicant used the CASMO5 code to determine the material concentrations of spent fuel and the MCNP code to determine the neutron multiplication factor, keff, of the package. The applicant performed code benchmarking analyses for the CASMO5 code using the radiochemistry assay (RCA) data from a collection of experiments as presented in Holtec report HI-2032982, Isotopic Benchmarks for Burnup Credit, Supplement 4, Revision

4. The applicant states that the NRC has accepted this code for burnup credit analyses in the previous applications. The staff verified that this statement is true and therefore did not perform any more detailed review of the suitability of the CASMO code for burnup credit analyses.

As discussed in Section 6.3.5 of this SER, the applicant used the direct difference method to determine the bias and bias uncertainty of the MCNP code for the major actinides as listed in Table 1 of ISG-8, Revision 3. This one of the approaches recommended in NUREG/CR-6811.

The applicant used HTC and MOX fuel with fresh water in its code benchmarking analysis for these isotopes. All trends are checked and the trend with the maximum bias (Pu concentration trend) is selected because it bounds all other trends. For this reason, this trend is therefore considered in the bias determination and used for the MPC-32M and F-32M fuel packages. The staff reviewed the applicants code benchmarking analyses for the actinides as listed in Table 1 of ISG-8, Revision 3. The staff finds that the direct difference method is appropriate for this benchmarking analyses because it is one of the recommended methods developed in NUREG/CR-6811.

For minor actinides and fission products (MAFPs), except Eu-151, as provided in Table 2 of ISG-8, Rev.3, the applicant used the recommendation of ISG-8, Rev. 3, which allows the applicant to use the bias (i) and bias uncertainty (ki) values estimated in NUREG/CR-7109 and NUREG/CR-7205 in lieu of explicit code benchmarking under some explicit conditions.

Specifically, ISG-8, Rev. 3 states:

In lieu of an explicit benchmarking analysis, the applicant may use the bias (i) and bias uncertainty (ki) values estimated in NUREG/CR-7108 using the Monte Carlo uncertainty sampling method, as shown in Tables 3 and 4 below. These values may be used directly, provided that:

• the applicant uses the same depletion code and cross section library as was used in NUREG/CR-7108 (SCALE/TRITON and the ENDF/B-V or -VII cross section library),

• the applicant can justify that its design is similar to the hypothetical GBC-32 system design used as the basis for the NUREG/CR-7108 isotopic depletion validation, and credit is limited to the specific nuclides listed in Tables 1 and 2,

• demonstrates that the credited minor actinide and fission product worth is no greater than 0.1 in keff.

Also, Table 5, Summary of code validation recommendations for isotopic depletion of ISG-8, Revision 3, explicitly requires that the applicant must demonstrate that its package design is similar to that of the GBC-32 cask. The table below is a copy of Table 5 of ISG-8, Revision 3.

It is important to note that the intent of the ISG for requiring the comparison of the applicants package design to compare the neutronic behavior of the two packages. Otherwise, the ISG will not be very useful because most of the cask designs are not identical to the GBC-32 cask which is used in the technical basis for the bias and bias uncertainty of the criticality safety analysis code.

Applicants Approach Recommendation Uses SCALE/TRITON and the ENDF/B- Use code bias and bias uncertainty V or -VII cross section library, and values from Tables 3 and 4 demonstrates that design application is similar to GBC-32

- or -

Uses other code and/or cross section Use either isotopic correction factor or library, or design application is not direct difference method to determine similar to GBC-32 code bias and bias uncertainty To use the recommendation, the applicant performed some analyses to compare the similarities between the GBC-32 cask and the HI-STAR 100MB package containing MPC-32M canister or the F-32M basket. The applicant concludes that overall, the evaluations show that the HI-STAR 100MB is sufficiently like the GBC-32 cask to justify the applicability of the method (Appendix 6.C of Holtec Report HI-2188084, Criticality Evaluation of HI-STAR 100MB).

Specifically, the applicant concludes: With respect to fuel type, fuel pitch, neutron absorber type, capacity, and cask design, the two systems are very similar. However, they differ in the basket material and amount of neutron absorber. The GBC-32 contains a steel basket with attached neutron poison plates, while the basket in the HI-STAR 100MB is made from an aluminum-based material with integrated neutron absorber material, at a significantly higher level. Due to the higher neutron poison level, the HI-STAR 100MB has a lower reactivity.

To see if that affects the neutronic behavior, the HI-STAR 100MB is analyzed with both its design basis neutron absorber, and with the same B-10 loading as the GBC-32 model. Table 6.C.2 shows the comparison of the hydrogen-to-fissile atom ratios (H/X), energy of average neutron lethargy causing fission (EALF), and the calculated keff of the systems. Calculations are performed for the MC-32M with design basis neutron absorber, the GBC-32, and the MPC-32M with the neutron absorber level adjusted to the same level as the GBC-32.

In addition, the applicant states:

• With respect to reactivity (calculated keff), the MPC-32M with design basis neutron absorber shows the lowest value, as expected. However, even when adjusted to the areal density value of the GBC-32, the reactivity is lower. This is because the neutron absorber is an integral part of the basket structure for the MPC-32M and therefore surrounds each assembly, while for the GBC-32 the neutron absorber is present in sheets and does not cover the corners of the cells. Again, it needs to be seen if these differences affect the more detailed neutronic behavior analyzed in the form of the neutron spectra and reaction rates.

• As expected, the EALF values follow this inversely, i.e. the higher reactivity basket has the lower EALF for the same fuel type and fuel composition.

• There are no significant differences between the different fuel compositions analyzed.

The staff reviewed the information presented in the application, including the results of the analyses and justifications. During its review, the staff finds that the CASMO computer code does not meet the conditions as quoted above to use the recommendation, i.e, taking a 1.5%

penalty of the reactivity worth of the MAFPs. In its letter dated February 4, 2019, the staff requested the applicant to demonstrate that the CASMO code meets the conditions for using the recommendation of ISG-8, Rev. 3 for treating the bias and bias uncertainty of the MAFPs.

In its response dated 3/14/2019 (ML19079A007), the applicant states: The overall approach addressing the MAFP bias is consistent with previous applications, namely for the HI-STAR 100 in Amendment 10 [1] and for the HI-STAR 190 [2]. Specifically, Appendix 6.C contains the comparison between the HI-STAR 100MB and the GBC-32, including consideration for two different assembly types in the HI-STAR 100 MB, and shows all those are neutronic equivalent.

Further Note that the MAFP bias has only a rather small effect on the overall evaluation. This bias is only of the order of 0.0015 delta-k, which is a small fraction of the overall bias and bias uncertainties (which is about 0.0246), and an even smaller fraction of the expected margin (more than 0.04 based on the best estimate calculation in Section 6.B.6).

The staff reviewed the SERs for both the HI-STAR 190 and HI-STAR 100 amendment 10. The staff finds that the SER for the HI-STAR 100 amendment 10 (ML18290A531) explicitly states that the approval is for the fuel from the Diablo Canyon reactor only. The staff did not perform a generic evaluation of the applicability of the method for other package/cask designs. Therefore, the approval of the cask design does not constitute a generic approval of the method of evaluation.

The HI-STAR 100 SER specifically states: The staff has reviewed the design description of and criticality analyses for the package. Based on its review of the application, the applicants responses to the staffs requests for additional information, the staff finds that the HI-STAR 100 package containing the MPC-68 or MPC-68F canister with revised thoria and UO2 content and enrichment, or the package containing MPC-32s canister for the Diablo Canyon Power Plant spent fuel, meets the regulatory requirements of 10 CFR 71.55 and 71.59. The staff also finds that the applicants use of the 1.5% of the reactivity of the MAFPs as the bias and bias uncertainty of the MCNP code to be acceptable for this specific application. The staff further verified that the application for and the Certificate of Compliance (CoC) of amendment 10 of the HI-STAR 100 cask are for the Diablo Canyon nuclear reactor spent fuel only. Based on these statements, the staff does not consider that the approval of HI-STAR 100 CoC amendment 10 established a basis for the approval of the HI-STAR 100MB.

Also, the staff finds that the SER for the HI-STAR 190 ((ML17222A084) does not explicitly states that the staff finds that the method used by the applicant for the MAFPs are generically acceptable. Specifically, the SER states: The applicant applied a bias of 1.5 percent of the worth of the minor actinides and fission products to account for the criticality code bias of these nuclides. This value was evaluated as an appropriate bias in NUREG/CR-7109, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety AnalysesCriticality (keff ) Predictions, for the SCALE code system for use with the ENDF/B-V, ENDF/B-VI, or ENDF/B-VII cross section libraries, and is found acceptable in accordance with the guidance in SFST-ISG-8, Rev. 3. NUREG/CR 7205, Bias Estimates Used in Lieu of Validation of Fission Products and Minor Actinides in MCNP Keff Calculations for PWR Burnup Credit Casks, has

extended this value for use with MCNP. Therefore, the staff finds it acceptable for use by the applicant for this application.

Based on these statements, it is clear that the staff approved the use of the recommendation of ISG-8, Rev. 3 for MAFPs was for the HI-STAR 190 only. On this basis, the staff does not consider that the approval of HI-STAR 190 established a basis for the approval method for use for the HI-STAR 100MB burnup credit analysis.

In addition, the staff finds that the code benchmarking analyses performed for HI-STAR 190 are not applicable to the HI-STAR 100MB because these two packages are significantly different in design; the HI-STAR 190 package has a capacity of 37 PWR fuel assemblies whereas the maximum capacity of the HI-STAR 100MB is 32 PWR fuel assemblies. Because of the differences in the geometric dimension, the leakage terms are significantly different, resulting in the significant differences in neutron importance in these two systems and hence the reactivity worth of the isotopes. These differences will affect the reactivity of the packages.

However, the staff finds that the SER for the HI-STAR 100 amendment 10 clearly states that it approved the design of the package based on that the reactivity worth of the Eu-151 is comparable with the penalty factor of 1.5% worth of the reactivity of the MAFPs as a penalty factor in lieu of explicit code benchmarking. Following the same logic, the staff evaluated the reactivity worth of the Eu-151 in the HI-STAR 100MB package containing the MPC-32M and F-32B and finds that the reactivity worth is comparable with that of the 1.5% reactivity of the bias and bias uncertainty of the MCNP code. On this basis, the staff finds that there is a reasonable assurance that omitting the credit of Eu-151 is likely to compensate the bias introduced in the CASMO code for not being able to track all key fission product like Eu-151.

It is important to point out that these depletion parameter values are directly taken from the input file of the CASMO depletion models provided by the applicant in its calculation package [Ref.

15]. But the codes lack of the ability to track all key fission products will skew the neutron flux and the absorption calculations of the isotopes; including fissile materials and absorbers that are being tracked. For these reasons, the staffs approval of this application is for this case only. It does not establish an endorsement of this approach because the code cannot guarantee that it will provide a consistent behavior for other fuel designs.

For the abovementioned reasons, the approval of this application (HI-STAR 100MB Revision 0) does not establish the basis for generic evaluation of the method. Future applications must use the recommendation of ISG-8, Rev. 3 in its entirety or be evaluated on a case by case basis.

6.9 Burnup verification method One of the requirements for applying burnup credit is to provide a means to verify that the burnup of the selected fuel meets the required minimum burnup assumed in burnup credit analyses. In the proposed the new method for burnup verification, the applicant implemented the recommendation of ISG-8, Rev. 3, performed a misloading analysis to demonstrate that even with accident misload, the cask remains subcritical. The applicant performed misload analyses for several scenarios and demonstrated the under these highly unlikely scenarios, the package remains subcritical. In the misloading analyses, the applicant used keff 0.98 as acceptance criterion.

The staff evaluated the method and results of the applicants misload analysis and finds it is consistent with the recommendation of ISG-8, Rev. 3, and therefore to be acceptable.

Specifically, ISG-8, Rev. 3 states: Misload analyses may be performed in lieu of a burnup measurement. A misload analysis should address potential events involving the placement of assemblies into a SNF storage or transportation system that do not meet the proposed loading criteria. The applicant should demonstrate that the system remains subcritical for misload conditions, including calculation biases, uncertainties and an appropriate administrative margin that is not less than 0.02 k.

The staff finds that the applicants analyses meet the acceptance criteria recommended by ISG-8, Rev. 3. On this basis, the staff finds the applicants burnup verification method to be acceptable.

6.10 Confirmatory Analysis The staff performed a confirmatory analysis for the most reactive configuration, i.e., water flooded and reflected single package containing the F-24M basket loaded with bounding 17x17 PWR assemblies. The staff use the SCALE 6.1 computer code with continuous energy cross sections derived from the ENDF/B-VII cross section library. The results confirm the applicants calculated keff value for the bounding package design.

6.11 Conclusions The staff reviewed the information provided in the application and the applicants responses to the staffs requests for additional information. Based on its review, the staff finds that the applicant made conservative assumptions in the criticality safety analyses, including maximum allowable quantity of fissile materials (assuming fresh fuel), conservative tolerance of cask geometry, reduced credit of B-10 in poison plates in the cask and the calculated maximum neutron multiplication factor, keff, with appropriate code benchmarking analyses.

Based on the review of the information presented by the applicant and its independent confirmatory analyses, the staff determined that the HI-STAR 100MB package meets the regulatory requirement of 10 CFR 71.55 and the acceptance criteria specified in NUREG-1617 on criticality safety with the following condition that no damaged fuel is authorized for transportation in this package.

6.12

References:

1. NUREG-1617, Standard Review Plan for Transportation Packages for Spent Nuclear Fuel, United States Nuclear Regulatory Commission, March 2000.

2. ANSI N14.5, American National Standard for Radioactive Materials Leakage Tests on Packages for Shipment", Institute for Nuclear Materials Management, American National Standards Institute, 2014.

3. Interim Staff Guidance - 19, Moderator Exclusion Under Hypothetical Accident Conditions and Demonstrating Subcriticality of Spent Fuel Under the Requirements of 10 CFR 71.55(e), United States Nuclear Regulatory Commission, May 2, 2003.

4. X-5 Monte Carlo Team, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5, LA-UR-03-1987, Los Alamos National Laboratory, April 2003 (Revised 2/1/2008),

5. ENDF/B-VII.1 Evaluated Nuclear Data Library, Brookhaven National Laboratory, December 22, 2011.

6. CASMO5/CASMO5M A Fuel Assembly Burnup Program Methodology Manual, SSP-08/405, Rev. 1, Studsvik Scandpower, Inc.

7. Interim Staff Guidance - 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transportation and Storage Casks, United States Nuclear Regulatory Commission, January 18, 2011.

8. Interim Staff Guidance - 23, Application of ASTM Standard Practice C1671-07 when performing technical reviews of spent fuel storage and transportation packaging licensing actions, United States Nuclear Regulatory Commission, September 26, 2012.

9. J.M. Cano, R. Caro, and J.M Martinez-Val, Supercriticality through Optimum Moderation in Nuclear Fuel Storage, Nucl. Technol., 48, 251-260, (1980).

10. J. J. Duderstandt, and L. J., Hamilton, Nuclear Reactor Analysis, John Wiley & Sons, Inc. 1976.

11. NUREG/CR-6811, Strategies for Application of Isotopic Uncertainties in Burnup Credit, Oak Ridge National Laboratory, June 2003.

12. NUREG/CR-7109, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety AnalysesCriticality (keff) Predictions, Oak Ridge National Laboratory, April 2012.

13. NUREG/CR-7205, Bias Estimates Used in Lieu of Validation of Fission Products and Minor Actinides in MCNP Keff Calculations for PWR Burnup Credit Casks, Oak Ridge National Laboratory, September 2015.

14. Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, Oak Ridge National Laboratory, June 2011

15. Holtec Report HI-2188084, Rev. 0, Criticality Analysis for HI-STAR 100MB and Input/Output Data Files, February 2018.

16. Holtec International Report HI-2012630, Burnup Credit for the MPC-32, March 23, 2004 (ML041450477).

17. Holtec Report No: HI-2156611, Burnup Credit for the MPC-32 with MCNP5 and CASMO5, January 29, 2016 (ML16173A235).

18. International Handbook of Evaluated Criticality Safety Benchmark Experiments, International Criticality Safety Benchmark Evaluation Project (ICSBEP),

https://www.oecd-nea.org/science/wpncs/icsbep/handbook.html 7.0 PACKAGE OPERATIONS

The package operations descriptions contain the essential elements of operations for using the package. Where the use of alternatives to described sequences or operations is acceptable, the operations descriptions include a description of these alternate sequences and operations.

The staff reviewed the applicants description of package operations to ensure (i) consistency with its technical evaluation, and (ii) compliance with the shielding design specified in the technical drawings and appropriate regulatory external dose rate limits. The staff finds that, based on its review, the operations descriptions in the application are consistent with the technical design and shielding analysis.

The staff reviewed the applicants description of package operations to ensure that they result in the package being used in accordance with the shielding design specified in the technical drawings and appropriate regulatory radiation limits. The applicant describes pre-loading inspections and package operations for both fuel in a MPC that has already been loaded as part of dry storage operations under 10 CFR Part 72 and fuel loaded directly into a bare basket. The staff reviewed the applicants operations and contents descriptions and found they are consistent with these considerations.

The applicant described the leak test requirements and operations in Section 7.1.4 of the application for a single lid cask closure (MPC), in Section 7.1.5 for preparation for transport (MPC), in Section 7.1.8 and Section 7.1.9 for dual lids cask closure (bare basket) and preparation for transport, respectively. The staff reviewed the leak test requirements and operations, as described, and finds them acceptable.

The applicant specified in Section 7.1.6 that an appropriate monitoring for combustible gas concentrations shall be performed along with providing the additional assurance that flammable gas concentrations will not be developed for loading the MPC with spent fuel. The applicant also specified in Section 7.1.7 that fuel exposure to air is not permitted under any scenario when loading and unloading the bare basket with spent fuel. The staff reviewed those Sections 7.1.6 and 7.1.7 and confirmed that such specifications are adequate for prevention of gas combustion during MPC loading and prevention of fuel exposure to air during bare basket loading and unloading. The applicant states, in Section 7.2.4, Removal of Contents from Bare Basket Cask, that: (a) gas sampling is performed to assess the condition of the fuel cladding. If a leak is discovered in the fuel cladding, the user's Radiation Control organization may require special actions to vent the cask cavity, and (b) an inert gas must be used any time the fuel is not covered with water to prevent oxidation of the fuel cladding. The fuel cladding is not to be exposed to air at any time during unloading operations. The staff reviewed Section 7.2.4 and finds that the use of gas sampling to assess the condition of the fuel cladding and the use of inert gas to prevent oxidation of the fuel cladding are acceptable methods.

The applicant stated in Section 7.1.5, Preparation for Transport, that the surface temperatures of the accessible areas of the package are measured to confirm the use or not of the personnel barrier if the accessible surfaces of the transport package exceed or not the exclusive use temperature limits of 49 CFR 173.442.

The staff reviewed Section 7.1.5 and finds that the measurement of the package surface temperatures is acceptable.

The applicant stated in Section 7.1.6, Loading the MPC with Spent Fuel, that (a) the combustible gas monitoring shall be performed prior to, and during MPC lid welding operations and the space under the MPC lid will be purged with an inert gas prior to, and during MPC lid

welding operations to provide additional assurance that flammable gas concentrations will not develop in this space and (b) the user should refer to Table 7.1.5 for load-and-go MPC drying limits and Table 7.1.4 for MPC backfill pressure requirements. The staff reviewed Section 7.1.6 and finds that the statements, operating conditions, and requirements for loading the MPC with spent fuel, are acceptable.

The applicant stated in Section 7.1.7, Loading the Bare Basket Cask with Spent Fuel, that for the bare basket with the spent fuel, (a) the package licensing and operating procedures do not permit fuel exposure to air under any scenario including loading and unloading operations, and (b) an inert gas must be used any time the fuel is not covered with water to prevent oxidation of the fuel cladding. The fuel cladding is not to be exposed to air at any time during loading operations, and (c) the user should refer to Table 7.1.6 of the application for bare basket cask drying criteria and Table 7.1.8 for bare basket cask backfill pressure requirements. The staff reviewed Section 7.1.7, and agrees with the statements, operating conditions, and requirements for loading the bare basket cask with spent fuel.

The applicant stated in Section 7.2, Package Unloading, that for MPC unloading and bare basket unloading, (a) the package licensing and operating procedures do not permit fuel exposure to air under any scenario including loading and unloading operations, and (b) an inert gas must be used any time the fuel is not covered with water to prevent oxidation of the fuel cladding. The fuel cladding is not to be exposed to air at any time during unloading operations.

The staff reviewed Section 7.1.5 and finds that the measurement of the package surface temperatures, as described in Section 7.1.5, is acceptable.

The applicant stated in Section 7.2.3, Removal of Contents from MPC, that combustible gas monitoring shall be performed prior to, and during. MPC lid welding operations and the space under the MPC lid shall be vented/exhausted or purged with an inert gas prior to, and during MPC lid cutting operations to provide additional assurance that flammable gas concentrations will not develop in this space. The staff reviewed Section 7.2.3 and finds that the combustible gas monitoring performed for MPC lid welding operations and MPC lid cutting operations -

during unloading of contents from the MPC - is acceptable.

The applicant stated in Section 7.2.4, Removal of Contents from Bare Basket Cask, that for removal of contents from bare basket cask, gas sampling is performed to assess the condition of the fuel cladding and special actions to vent the package cavity may be required if a leak is discovered in the fuel cladding. The applicant also stated in that same Section that, for the bare basket cask, an inert gas must be used any time of unloading of contents when the fuel is not covered with water to prevent oxidation of the fuel cladding. The fuel cladding is not to be exposed to air at any time during unloading operations.

The staff reviewed the Operating Procedures in Chapter 7 of the application to verify that the package will be operated in a manner that is consistent with its design evaluation. On the basis of its evaluation, the staff concludes that the combination of the engineered safety features and the operating procedures provide adequate measures and reasonable assurance for safe operation of the proposed design basis fuel in accordance with 10 CFR Part 71.

Further, the Certificate of Compliance states that the package must be prepared for shipment and operated in accordance with the Operating Procedures specified in Chapter 7 of the application and includes also additional conditions stemming from the staffs technical review.

8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM

Chapter 8 of the application identifies the inspections, acceptance tests and maintenance programs to be conducted on the Model No. HI-STAR 100MB package and verifies their compliance with the requirements of 10 CFR Part 71.

The package shielding design includes the packages steel shells, lids, and base. It also includes the radial lead shielding and neutron shield assemblies. The shielding effectiveness is ensured in part by confirming the package components are fabricated to the specifications, including tolerances described in the design drawings that are part of the CoC. It is also ensured in part by confirming package components do not have any defects and that the package and its components are fabricated and assembled properly. The applicant describes the method to verify the adequacy of the radiation shielding in Section 8.1.5.7 of the application.

Staff reviewed the acceptance tests described in Section 8.1.5 and finds that these tests are adequate to ensure shielding effectiveness.

The applicant also includes descriptions of the package maintenance in Section 8.2 of the application. While the packages gamma and neutron shielding are not expected to degrade over time, the applicant included a test in Section 8.2.2.2 of the application to ensure continued effectiveness of the packages shielding over the life time of the package. This test is to be conducted within five years prior to shipment. The staff reviewed the applicants maintenance programs and finds it acceptable based on the acceptance criteria provided in NUREG-1617.

The applicant presented the containment system performance specifications in Table 8.1.1 of the application. Table 8.1.2.a provides a list of the systems (MPC and cask) and components that are tested, the type of leakage rate test and their allowable leakage rates for (i) the fabrication leakage rate test, (ii) the pre-shipment leakage rate test, (iii) the maintenance leakage rate test and (iv) the periodic leakage rate test for single lid cask containment systems.

It is also noted, in Table 8.1.2.a, that the leakage testing of the MPC (MPC-32M) is required if it contains high burnup fuel.

The applicant also defined in Table 8.1.2.b the system (cask) and components being tested, the type of leakage rate test and allowable leakage rate for (i) the fabrication leakage rate test, the pre-shipment leakage rate test, (ii) the maintenance leakage rate test and (iii) the periodic leakage rate test for dual lid cask containment systems. The staff reviewed those two Tables and confirmed that the leakage rate tests that are specified are consistent with the leakage rate tests presented in Sections 4.1 and 4.1.4 for the single lid cask (MPC) and Sections 4.1 and 4.1.5 for the dual lid cask (bare basket).

Section 8.1.5.8, Thermal Test, specifies that the first fabricated HI-STAR 100MB package shall be tested to confirm its heat dissipation capability. The thermal test is considered acceptable if the measured heat rejection capability is greater than the design basis minimum heat rejection capacity.

The applicant stated in Section 8.1.6.2, Post-Shipment HBF Integrity Acceptance Test, that for packages containing HBF, package surface temperatures and package shall be measured in accordance with procedures in Chapter 7 as a practical means of monitoring the condition of the fuel assemblies. The staff reviewed Section 8.1.6.2 and finds the use of the post-shipment Integrity acceptance test to monitor the condition of the HBF to be acceptable.

MPC Surface Defect Inspection

As described in Sections 1.2.1.2 and 8.1.6 of the application, due to uncertainty in fuel cladding properties, an MPC containing high burnup fuel is credited as a second containment barrier to prevent the intrusion of water.

To demonstrate the integrity of the containment boundary during transportation, all MPCs containing high burnup fuel are leakage tested, and, as discussed below, the containment boundaries of a sample of older MPCs are inspected prior to shipment to ensure that no flaws exist that could compromise containment under HAC.

Within one year prior to package shipment, all MPCs are leakage tested in accordance with ANSI N14.5-2014 to a rating of 1x10-7 cm3/s air. Also, for MPCs stored at an ISFSI for more than 5 years under 10 CFR Part 72, all MPC shells are inspected visually with a remote camera and, on a sampling basis, with eddy current techniques to verify the absence of age-related surface defects with greater than 2 mm depth.

The applicant provided a proprietary evaluation to demonstrate that an MPC with the maximum-allowable 2 mm deep flaw would be able to sustain HAC loads. Any flaw detected on the MPC surface that exceeds the maximum-allowable depth shall not be accepted for transport.

Table 8.1.6 summarizes the sampling approach used to perform eddy current examinations of MPCs containing high burnup fuel. The approach starts with testing of 20 percent of a lot of MPCs, focusing on the lead canisters (those considered most susceptible to aging). A lot size is defined by the CoC user as some fraction of MPCs at an ISFSI and can range from one MPC to all the MPCs at the site. After a number of lots pass the eddy current testing with no defects, the percentage of MPCs tested in subsequent lots can be reduced (i.e., the sampling percentage is moved down a tier).

If any MPC fails an eddy current test, then every MPC in that lot must be tested and the percentage of MPCs in subsequent lots must be increased (i.e., sampling is moved back up a tier). The staff notes that this approach can lead to some variation in the number of MPCs tested depending on a users definition of a lot size and the success as passing the tests; however, in practice the sampling size would be a minimum of approximately 20 percent of the MPCs exceeding 5 years of storage at an ISFSI storing 50 MPCs.

The staff evaluated the proposed acceptance tests for canisters containing high burnup fuel and finds them to be acceptable because the combination of leaking testing of all canisters, visual examinations of the entire surface of all canisters, and volumetric (eddy current) examinations of a sample of canisters most susceptible to degradation is considered adequate to ensure that the canisters will be capable to excluding water in a transportation accident.

Based on the statements and representations in the application, the staff concludes that the acceptance tests for the packaging meet the requirements of 10 CFR Part 71. Further, the Certificate of Compliance specifies that each package must meet the Acceptance Tests and Maintenance Program of Chapter 8 of the application while including also the conditions described above.

CONDITIONS The following conditions are included in the Certificate of Compliance:

(a) The package shall be prepared for shipment and operated in accordance with the Operating Procedures in Chapter 7 of the application

(b) The package must meet the Acceptance Tests and Maintenance Program of Chapter 8.0 of the application.

(c) Damaged fuel assemblies, fuel debris, and irradiated non-fuel hardware are not authorized for transportation.

(d) Maximum allowable time for the completion of wet transfer operations, based on design basis maximum heat load and initial pool water temperature of 48.9°C, is 17.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

The maximum allowable time maybe recalculated, with other cask heat loads and pool water measured temperatures, prior to loading operations, as specified in Section 7.1.7.3 of the application.

(e) The vacuum drying operations do not prescribe time limits for (i) the F-24M and F-32 M baskets respectively, for high burnup fuel, provided the cask heat load is equal to or below 24 kW and 26 kW respectively, and (ii) the MPC-32M for both high burnup fuel and all configurations with moderate burnup fuel.

(f) The minimum specific power, the maximum moderator temperature, the maximum fuel temperature of each fuel design shall not exceed the values listed in Table 7.7.4 of the application.

(g) The fuel burnup credit loading curve is applicable only to spent fuel assembly classes, loaded in the MPC-32M/F-32M, identified in Table 7.7.3(a) of the application.

(h) The package shall be transported exclusive use only with the personnel barrier installed during transport.

(i) Transport of fissile material by air is not authorized.

CONCLUSION Based on the statements and representations contained in the application, and the conditions listed above, the staff concludes that the Model No. HI-STAR 100MB 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. 9378, Revision No. 0, On August 9, 2019.