ML20174A555

Draft SER

ML20174A555
Person / Time
Site:	07201031
Issue date:	07/28/2020
From:	Office of Nuclear Material Safety and Safeguards
To:	NAC International
BWhite NMSS/DFM/STL 415.6577
Shared Package
ML20174A550	List: ML20174A551 ML20174A552 ML20174A553 ML20174A554 ML20174A555
References
Download: ML20174A555 (27)

Text

Enclosure 3 PRELIMINARY SAFETY EVALUATION REPORT NAC INTERNATIONAL, INC.

MAGNASTOR STORAGE SYSTEM DOCKET NO. 72-1031 AMENDMENT NO. 9 Summary By application dated October 9, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19296C938), as supplemented on April 9, 2020 (ADAMS Accession No. ML20108F317), and June 29 (ADAMS Accession No. ML20192A118), NAC International (NAC) submitted an application for Amendment No. 9 to Certificate of Compliance (CoC) No. 1031. The applicant requested addition of a new concrete overpack (named CC6);

four new heat load zone patterns; a new hybrid fuel assembly type (called BW15H5); a new maximum enrichment for the BW15H2 hybrid fuel assembly, along with a new minimum soluble boron concentration and neutron absorber areal density; and reorganization of Appendix B of the technical specifications (TSs). This safety evaluation report (SER) documents the U.S.

Nuclear Regulatory Commission (NRC) staffs review and evaluation of Amendment No. 9 for CoC No. 1031 for the Modular Advanced Generation Nuclear All-purpose STORage (MAGNASTOR) system.

The NRC staff reviewed the amendment request using guidance in NUREG-1536, Revision 1, Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility, Final Report, Rev. 1, dated July 2010. For the reasons stated below and based on the statements and representations in NACs application, as supplemented, and the conditions specified in the CoC and TSs, the staff concludes that the requested changes meet the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste.

Chapter 1 GENERAL INFORMATION EVALUATION The objective of the review of this chapter is evaluate design changes made to the MAGNASTOR storage system to ensure that NAC provided a description that is adequate to familiarize reviewers and other interested parties with the pertinent features of the system, including the requested changes.

1.1 General Description and Operational Features The MAGNASTOR system is a spent fuel, dry storage system consisting of a concrete cask and a welded stainless steel canister (the transportable storage canister, or TSC) with a welded closure to safely store spent fuel. In the storage configuration, the TSC is placed in the central cavity of the concrete cask. The concrete cask provides structural protection, radiation shielding, and internal airflow paths that remove the decay heat from the TSC surface by natural air circulation. The concrete cask also provides protection during storage for the TSC against adverse environmental conditions. The system is designed to accommodate storage of pressurized-water reactor (PWR) and boiling-water reactor (BWR) spent fuel. The MAGNASTOR system is designed to store up to 37 PWR or up to 87 BWR spent fuel assemblies in each TSC in the appropriate fuel basket type. In addition to the TSC and the concrete cask, the other principal component of the MAGNASTOR system is the transfer cask.

The transfer cask is used to move the TSC between the workstations during TSC loading and preparation activities, and to transfer the TSC to or from the concrete cask.

1.1.1 Storage Overpack NAC proposed adding Concrete Cask Number 6 (CC6) storage overpack, which is a 210.1-inch tall, 136-inch diameter cylinder. CC6 includes a 3-inch thick carbon steel liner. The concrete shell is 25.3 inches thick with rebar of various lengths. The diameter of the lid is 91.5 inches and its thickness is 12.3 inches. CC6 is equipped with additional shielding at the air inlets. The upper air outlets of CC6 are incorporated into the lid.

1.1.2 Transportable Storage Canister The TSC is unchanged, however there were minor changes to the basket that do not affect the performance of the basket, such as accounting for changes made under the authority of 10 CFR 72.48, reducing the tolerance by 0.01 inches on tube thickness, removing duplicative dimensions, and removed changes from a withdrawn amendment. The width of the four cells for damaged fuel in the PWR damaged fuel (PWR-DF) basket opening may be slightly increased for the newly added hybrid fuel assembly, named BW15H5. The damaged fuel can (DFC) which is fabricated from Type 304 stainless steel and has an 8.7 to 9-inch-square inside dimension fits inside the four cells in the PWR-DF basket.

1.1.3 Transfer Cask The transfer casks are unchanged, and transfer of the new BW15H5 fuel associated with this amendment is required to use the MAGNASTOR transfer cask number 2 (MTC2), which is a shortened stainless steel version of the original carbon steel transfer cask, which is designated MTC in the TSs. The certificate of compliance was clarified to state that that the MTC2 is a stainless steel version of the original MTC.

1.2 Drawings In support of this application, NAC submitted the following 9 proprietary drawings for NRC review:

Drawing No. 71160-551, Rev. No. 13P - Fuel Tube Assembly, MAGNASTOR - 37 PWR Drawing No. 71160-575, Rev. No. 13P - Basket Assembly, MAGNASTOR - 37 PWR Drawing No. 71160-601, Rev. No. 3P - Damaged Fuel Can (DFC), Assembly, MAGNASTOR Drawing No. 71160-602, Rev. No. 4P - Damaged Fuel Can (DFC), Details, MAGNASTOR Drawing No. 71160-661, Rev. No. 0P - Structure, Weldment, - Concrete Cask, MAGNASTOR Drawing No. 71160-662, Rev. No. 0P - Reinforcing Bar and Concrete Placement, Concrete Cask, MAGNASTOR Drawing No. 71160-663, Rev. No. 0P - Lift Lug and Details, Concrete Cask, MAGNASTOR Drawing No. 71160-664, Rev. No. 0P - Upper Segment Assembly, Concrete Cask, MAGNASTOR Drawing No. 71160-690, Rev. No. 0P - Loaded Concrete Cask, MAGNASTOR 1.3 Contents The contents have been revised to add four new heat load zone patterns and associated decay heats that are specific to Babcock and Wilcox (B&W) 15x15 fuel assemblies as shown in Tables B2-2 and B2-8 of Appendix B of the TSs for use with the new CC6 and the MTC2. NAC proposed adding a new hybrid fuel assembly type, BW15H5, and its associated fuel assembly characteristics. NAC also proposed increasing the maximum enrichment to 5.0 weight percent (wt%) 235U for the BW15H2 fuel assembly hybrid. For the increased enrichment, the hybrid BW15H2 assembly requires a new minimum soluble boron concentration of 2650 parts per million (ppm) of boron-10 (10B) and neutron absorber panels that that have a minimum areal density of 0.036 grams of 10B per cubic centimeter (g/cm3).

1.4 Evaluation Findings

Based on the NRC staff's review of information provided for the Amendment No. 9 to the MAGNASTOR system, the staff determined the following:

F1.1 A general description and discussion of Amendment No. 9 to MAGNASTOR system is presented in Chapter 1 of the SAR, with special attention to design and operating characteristics, unusual or novel design features, and principal safety considerations, and the description is sufficient to familiarize a reviewer or stakeholder with the design.

F1.2 Drawings for structures, systems, and components (SSCs) important to safety presented in Section 1.8 of the SAR were reviewed. Details of specific SSCs are evaluated in Sections 3 through 12 of this SER.

Chapter 2 PRINCIPAL DESIGN CRITERIA EVALUATION The changes associated with principal design criteria include addition of the new CC6, the four preferential heat load patterns and their associated decay heats, addition of the BW15H5 hybrid fuel assembly type and reconfiguration of the TS. These changes are discussed and evaluated in subsequent chapters of this SER.

Chapter 3 STRUCTURAL EVALUATION The staff reviewed the proposed changes to verify that the applicant has performed adequate structural evaluation to demonstrate that the system, as proposed, is acceptable under normal, off-normal, accident conditions, and natural phenomena events. In conducting this evaluation, the staff seeks reasonable assurance that the system will maintain confinement, subcriticality, radiation shielding, and retrievability or recovery of the fuel, as applicable, under all credible loads of normal, off-normal, accident conditions, and natural phenomenon events.

The following proposed changes are applicable to the structural evaluation:

addition of CC6 clarifying (non-technical) changes to TSs, Appendix B 3.1 Addition of a new Concrete Overpack 3.1.1 Design Description of Concrete Cask Number 6 The CC6 is a reinforced concrete cask, which has a similar design to the first concrete overpack (originally named CC, but designated as CC1, and used here throughout) for which the staff previously reviewed and approved in CoC No. 1031, Amendment No. 0 (ADAMS Accession No. ML090350509). It has dimensions of an outside diameter of 136 inches and an overall height of approximately 210 inches. The internal cavity of the concrete cask is lined by a carbon steel liner which has an inside diameter of 79.5 inches and a thickness of 3.0 inches. The concrete shell, constructed using Type II Portland Cement, has a nominal density of 145 pounds per cubic foot and a compressive strength of 8,000 pounds per square inch (psi) at ambient temperature.

A ventilation airflow path is formed by inlets at the bottom of the concrete cask, the annular space between the concrete cask inner shell and the TSC, and outlets in the concrete cask lid assembly. The passive ventilation system operates by natural convection as cool air enters the bottom inlets, is heated by the TSC, and exits from the outlets. Both the air inlets and air outlets are formed with carbon steel in the concrete cask body.

3.1.2 Design Criteria of Concrete Cask Number 6 The applicant primarily used the structural design criteria described in Chapter 2, Principal Design Criteria, of the final safety analysis report (FSAR) (ADAMS Accession No. ML14176B275), which were previously reviewed and approved by the staff, for the design of the CC1, with the exception of the concrete compressive strength. The CC6 is a reinforced concrete structure that is designed in accordance with the requirements of American Concrete Institute (ACI) 349, Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary, and the American Institute of Steel Construction (AISC) Manual of Steel Construction. The staff reviewed the design criteria for CC6 and found it to be consistent with the ACI 349 and AISC codes and standards.349 and AISC codes and standards.

3.1.3 Design Load Combinations of Concrete Cask Number 6 The load combinations used for the evaluation of the CC6 are identical to the load combinations previously used for the evaluation of the CC1 described in the FSAR (ADAMS Accession No. ML14176B275). Table 3.1 provides a summary of the seven load combinations used for the structural evaluations of the CC6. The staff reviewed the load combinations and found that they are consistent with American National Standards Institute (ANSI) 57.9, Design Criteria for an Independent Spent Fuel Storage Installation (Dry Storage Type, and therefore are acceptable.

Table 3.1 - Load Combination for CC6 Component Evaluation Load Combination Event 1.4 DL + 1.7 LL Normal 1.05 DL + 1.275 (LL + To)

Off-Normal 1.05 DL + 1.275 (LL + To + W)

Off-Normal - Wind DL + LL + To + E Accident - Earthquake DL + LL + To + Wt Accident - Tornado DL + LL + To + FL Accident - Flood DL + LL + Ta Accident - Thermal DL-Dead Load LL - Live Load To - Normal Temperature W - Wind Wt - Tornado/Tornado Missile E - Design Basis Earthquake FL - Flood Ta - Off-Normal or Accident Temperature 3.1.4 Structural Evaluation for Concrete Overpack Number 6 Lift Analysis of CC6 The CC6 is lifted using two lift lug assemblies that bolt to the upper forging of the cask liner weldment. A lift ring is welded to the outside of the cask liner and embedded in the concrete near the bottom of the cask to provide an anchor in the concrete for lift. The applicant evaluated the lift lug assembly and the lift ring including the lift of the cask upper segment.

The applicant used a combination of finite element analysis (FEA) using ANSYS computer software program and hand calculations to evaluate the concrete cask lift. This analysis method is the same method used for the evaluations of CC1, which the staff previously reviewed and approved in CoC No. 1031, Revision No. 0.

Table 3.2 provides results of the structural analysis of the CC6. The staff reviewed the stress calculations and found that the calculated stresses induced by the lifting operations are less than the allowable stresses, and thereby the calculated factor of safety (FS) based on the yield strength is larger than the FS specified in ANSI N14.6, Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10 000 Pounds (4500 kg) or More, for each component, which indicates that the design of the concrete cask for lift is acceptable.

Table 3.2 - Evaluation of Concrete Cask for Lift Component Calculated Value Calculated Factor of Safety (FS)

ANSI N14.6 Factor of Safety (FS)

Lift Lug 10.6 ksi*

4.0 3.0 Embedded Lift Ring Weldment 4.62 ksi 4.6 1.0 Concrete Bearing 0.8 ksi 5.4 1.0 Concrete Shear 0.04 ksi 130.0 1.0 Lift Pin 679 kips 3.4 3.0 Lift Lug Bolt 20.3 ksi 4.4 3.0 Lift Lug Weldment 8.8 ksi 5.5 3.0 Upper Segment 4.0 ksi 8.2 6.0

• kilopound per square inch (ksi)

Structural Analysis of CC6 for Normal and Off-Normal Conditions The applicant calculated the maximum stress of 1,424 psi on the inner surface of the CC6 using the ANSYS FEA computer software program when the load combination Number 3 from the FSAR [1.05 DL + 1.275 (LL + To + W)] was applied to the cask for the normal and off-normal conditions. Since the allowable compressive stress for the CC6 concrete cask is 5,040 psi, the calculated minimum FS for the normal and off-normal conditions is 3.54. Because the calculated FS of 3.54 is greater than the ANSI N14.6 FS of 1.0, the staff finds the performance of the CC6 against the normal and off-normal conditions is acceptable.

Structural Analysis of CC6 for Accident Conditions Using the ANSYS FEA computer software program, the applicant calculated the maximum stress of 1,482 psi on the inner surface of the CC6 for the accident conditions when the load combination Number 5 from the FSAR (DL + LL + To + Ess) was applied. Since the allowable compressive stress for the CC6 concrete cask is 5,040 psi, the calculated minimum FS for the accident conditions is 3.4. Because the calculated FS of 3.2 is greater than the ANSI N14.6 FS of 1.0, the staff finds the performance of the CC6 against the accident conditions is acceptable.

Stability Analysis of CC6 for Tornado and Tornado-Generated Missile The applicant performed an overturning analysis of the CC6 under tornado wind loading. The applicant used the same analytical approach, which was previously reviewed and accepted by the staff for the overturning analyses in the FSAR for CC1, which the staff previously reviewed and approved in CoC No. 1031, Revision No. 0. The applicant considered the maximum wind pressure, gust factor and cask dimensions, and calculated overturning moment of 4.4 x 105 foot-pounds (ft-lb). It also calculated the stability moment of 1.13 x 106 ft-lb for the cask and calculated the minimum FS against overturning as FS=1.72 based on the American Society of Civil Engineers (ASCE) 7-93, Minimum Design Loads for Buildings and Other Structures,. The staff finds the evaluation of the CC6 against the tornado wind loading is acceptable.

Additionally, the applicant also calculated a penetration depth of a tornado wind-generated missile for the CC6 concrete shell. The applicant used the same analytical approach, which was previously used for the CC1, which the staff previously reviewed and approved in CoC No.

1031, Revision No. 0. After consideration of the missile characteristic (e.g., velocity, shape, weight, diameter, shape factor) and cask dimensions, the applicant calculated a penetration depth of 4.96 inches. Since minimum concrete shell thickness to prevent scabbing is three times the penetration depth (3 x 4.96 inches = 14.9 inches) and the minimum thickness of the concrete shell is 25.2 inches, the FS = 25.2/14.9 = 1.69, which is larger than the ANSI N14.6 FS of 1.0. The staff finds the calculation for the penetration of the wind-generated missile into the CC6 concrete shell is acceptable.

Stability Analysis of CC6 for Flood The applicant performed an overturning analysis of the CC6 under a design-basis flood accident. The applicant used the same analytical approach, which was previously reviewed and approved by the staff for the overturning analyses of the CC1, which the staff previously reviewed and approved in CoC No. 1031, Revision No. 0. The applicant considered the factors (i.e., drag force of the flood, cask dimensions, etc.) and calculated a water velocity of 19.8 feet per second (ft/sec) required to overturn the cask, which provides a FS 1.32 against the flood loading. The staff reviewed the applicants calculation and finds the evaluation of the CC6 against the flood loading is acceptable.

Stability Analysis of CC6 for Earthquake The applicant performed an overturning analysis of the CC6 under an earthquake accident. The applicant used the analytical approach, which is presented in Section 3.7.3.4 of the FSAR, Amendment No. 0 and was previously reviewed and accepted by the staff for the overturning analyses of the CC1, which the staff previously reviewed and approved in CoC No. 1031, Revision No. 0. The applicant calculated an acceleration of 0.407g, which is an acceleration of the CC6 to resist an overturning. The applicant evaluated a minimum earthquake ground acceleration of 0.25g (which is greater than the minimum design-basis earthquake in 10 CFR 72.102(a) for specifically licensed ISFSIs). Therefore, the minimum FS against overturning of the CC6 under earthquake loading is FS = 0.407g/0.25g = 1.63, which is larger than the required FS of 1.1. The staff reviewed the applicants evaluation of the CC6 against the earthquake loading and finds the evaluation acceptable.

24-inch Drop Analysis for CC6 The applicant calculated a crush depth of the CC6 concrete cask for the 24-inch drop using an energy balance equation. The drop height, cross-sectional area, weight and compressive strength of the concrete cask were considered in the equation, and a crush depth of 0.06 inch was calculated. The applicant did not further evaluate the 24-inch drop analysis for the CC6 because this crush depth of 0.06 inch of the CC6 is less than the crush depth for the CC1, which was previously reviewed and approved by the staff, and therefore the evaluation of the CC6 is bounded by the evaluations of CC1. The staff finds the applicants statement acceptable.

Structural Analysis for CC6 for Tip-Over The applicant performed an evaluation of the CC6 under a non-mechanistic tip-over condition as a defense-in-depth measure and a continuation of its original certification basis in CoC No. 1031, Amendment No. 0. In order to demonstrate adequate performance of the CC6 for a tip-over event, the applicant first calculated an angular velocity, 6, of the CC6 with an assumption that the cask behaves as a rigid body. The angular velocity value was determined using conservation of energy (potential energy equals kinetic energy of the system, mgh = I.2/2) about the center-of-gravity (CG) as the concrete cask rotates from end to corner to side orientations. This is the same approach that was taken for CC1, where the angular velocity () of CC1 was used as an input into the LS-DYNA model that evaluated the tip-over event in depth.

The applicant calculated a ratio of the angular velocity of the CC6, 6, with respect to the angular velocity of the CC1, 1. The calculated ratio (6/1) was about 1.02. Based on this ratio of 1.02, the applicant concluded that since the ratio is very small, no tip-over analysis is required for the CC6 because, given the similar angular velocity and geometry between the CC6 and CC1, it is appropriate to use the results of the tip-over analysis presented in the FSAR, Revision 0, MAGNASTOR for CC1.

The staff reviewed the applicants approach and statement and although the staff finds that use of angular velocity alone is a simplistic approach that does not consider the intricacies of complex impact problems involved in evaluating the non-mechanistic tip-over. A more comprehensive evaluation that can evaluate the non-linearity in the analysis (e.g., an analytical evaluation with a LS-DYNA model) should be used for analyses involving more significant changes in angular velocity than that used for CC6 when compared to the tip-over analysis for CC1. Specifically, a comprehensive tip-over analysis should also consider variations in: (i) material properties (soils underneath the independent spent fuel storage facility (ISFSI) pad and the concrete pad [if different from the pad/soil used for CC1]; concrete cask; steel liner; basket; and fuel), (ii) geometric and material non-linear behavior during time-dependent dynamic impact loading, and (iii) interactions between soil-pad, pad-cask, and cask-internals (liner, basket and fuels), in addition to an angular velocity. This is the approach used for the analysis of tip-over for the CC1. Angular velocity is only one parameter among many modeling parameters that should be considered in the analyses of a tip-over event.

However, despite having not performed a more in-depth tip-over analysis, the staff has concluded that no additional non-mechanistic tip-over analysis of the CC6 is needed, in this instance, because there is reasonable assurance that the CC6 will perform its intended safety functions under a non-mechanistic tip-over event. This is due to conservatism and similarity of the CC6 to other applicants concrete casks as shown in Table 3.3 of this SER below.

Specifically: (i) CC1 was designed with an additional 50% margin with the g-loads calculated by LS-DYNA tip-over analysis (i.e., design-basis of 35.0g and 40.0g at the top of the fuel basket and cask, respectively, compared to calculated g-loads); (ii) both the CC1 and CC6 are evaluated on the same pad; (iii) the CC1 and CC6 are of similar construction; (iv) the CC6 is shorter and has a slightly shorter center of gravity as compared to the CC1, therefore it is more stable; and (v) the initial angular velocity of CC6 is within 2% of the CC1.

Table 3.3 - g-load at Top of the Fuel Basket and Cask Cask Type Method Fuel Basket (Design Basis = 35g)

Cask Design Basis = 40g CC1 LS-DYNA 26.4g 29.5g 3.2 Revision to Technical Specification, Appendix B The staff reviewed the proposed changes to the TS in Appendix B and finds that the changes are non-technical and Appendix B to the TS accurately reflects the addition of the CC6 to the MAGNASTOR cask system.

3.3 Evaluation Findings

F3.1 On the basis of the review of the statements and representations in the application, the staff finds that the applicant adequately describes CC6 in Section 3.11 Structural Evaluation for CC6 Concrete Cask to enable an evaluation of its structural performance and effectiveness.

F3.2 The staff finds that the applicant has met the requirements of 10 CFR 72.236(b). The CC6 is designed to accommodate the combined loads of normal, off-normal and accident and natural phenomena events with an adequate margin of safety. Stresses at various locations of the cask for various design loads are determined by analysis. Total stresses for the combined loads of normal, off-normal, accident, and natural phenomena events are acceptable and are found to be within the limits given in the applicable codes, standards, and specifications.

The staff concludes that the structural performance of CC6 that comprises the MAGNASTOR cask system is in compliance with 10 CFR Part 72, and that the applicable design and acceptance criteria in Section 3.4 of NUREG-1536, have been satisfied. The evaluation of structural performance provides reasonable assurance that the MAGNASTOR cask system will allow for the safe storage of spent fuel for the licensed period. This finding is reached on the basis of a review that considered the regulation itself, appropriate regulatory guides, applicable codes and standards, and accepted engineering practices.

Chapter 4 THERMAL EVALUATION The thermal review of Amendment No. 9 for the MAGNASTOR cask system ensures that the cask components and fuel material temperatures will remain within the allowable values under normal, off-normal, and accident conditions. This review includes confirmation that the fuel clad temperatures for fuel assemblies stored in the MAGNASTOR cask system will be maintained below specified limits throughout the storage period in order to protect the cladding against degradation that could lead to gross ruptures. This portion of the review also confirms that the cask thermal design has been evaluated using acceptable analytical techniques and/or testing methods.

This review was conducted under the regulations described in 10 CFR 72.236, which identify the specific requirements for the regulatory approval, fabrication, and operation of spent fuel storage cask designs. The unique characteristics of the spent fuel to be stored in the MAGNASTOR cask system are identified, as required by 10 CFR 72.236(a), so that the design basis and the design criteria that must be provided for the SSCs important to safety can be assessed under the requirements of 10 CFR 72.236(b).

This application was also reviewed to determine whether the MAGNASTOR design fulfills the acceptance criteria listed in Sections 2, 4 and 12 of NUREG-1536.

The following changes proposed under Amendment No. 9 to the MAGNASTOR cask system are applicable to the thermal evaluation:

Adding CC6.

Adding four new zoned loading patterns and associated maximum heat loads that are specific to Babcock and 'Wilcox (B&W) 15x15 fuel assemblies. These zoned loading patterns are only authorized for use in CC6 and the MTC2.

4.1 MAGNASTOR System Thermal Model The applicant used the ANSYS FLUENT computer-based analysis program to evaluate the thermal performance of the MAGNASTOR spent fuel storage system. ANSYS FLUENT is a finite volume computational fluid dynamics (CFD) program with capabilities to predict fluid flow and heat transfer phenomena in two and three dimensions. Chapter 1 of the SAR provides a general description of CC6 and SAR Section 4.11.1 provides a general description of the thermal model. The applicant developed a three-dimensional (3-D) quarter-symmetry finite volume model to evaluate the thermal performance of the CC6 and TSC for the 15x15 PWR fuel configuration. The thermal model includes concrete cask (including lid, liner and pedestal top); air in the air inlets; annulus with stand-offs and the air outlets; TSC shell, lid and bottom plate; fuel basket; fuel assemblies (represented as porous media); and helium internal to the TSC. Boundary conditions and air flow characteristics were identical to previous analysis, as described in FSAR Chapter 4, which the staff previously reviewed and approved in CoC No.

1031, Revision No. 0. For ambient temperature the applicant assumed 76°F. The staff noticed this temperature seems to be nonconservative for sites where ambient temperature would be higher (including seasonal variations). For this case, a general licensee would need to demonstrate the cask thermal analysis assuming 76°F bounds the site. Otherwise, site-specific analysis would need to be performed based on the site characteristics (using a maximum normal ambient temperature) to provide bonding results.

The staff reviewed the applicants description of the MAGNASTOR storage system thermal model. Based on the information provided in the application regarding the thermal model, the staff determined that the application is consistent with guidance provided in NUREG-1536, Section 4.4.4 (Analytical Methods, Models, and Calculations). Therefore, the staff concludes that the description of the thermal model is acceptable, as the description is consistent with NUREG-1536, and satisfies the requirements of 10 CFR 72.236(b), 10 CFR 72.236(f), 10 CFR 72.236(g), and 10 CFR 72.236(h).

4.2 Thermal Evaluation for Normal Conditions of Storage The applicant used the 3-D thermal model, described in Section 4.11.1 of the SAR, to determine temperature distributions under long-term normal storage conditions. The applicant performed thermal calculations for the four heat load patterns, X, Y, Z, and Z-PRIME, shown in Figures 4.11-1 through 4.11-4 of the SAR. The maximum fuel cladding temperature was obtained for heat load pattern Z, as shown in Table 4.11.1 of the FSAR. All predicted temperatures (including maximum fuel cladding temperature) remain below the allowable limits provided in the SAR.

For the bounding configuration (heat load pattern Z), the maximum average helium temperature in the TSC is bounded by the maximum average helium temperature calculated in Revision 0 of the FSAR Section 4.4.4 for the internal pressure calculation. Therefore, the maximum normal condition pressure for the TSC containing the preferential loaded B&W 15x15 fuel is bounded by the maximum normal condition pressure calculated in FSAR Section 4.4.4.

The staff reviewed the applicants thermal evaluation of the MAGNASTOR storage system during normal conditions of storage for the addition of CC6 and heat load patterns to the TSC canister for PWR fuel. Based on the information provided in the application regarding the thermal model and the evaluation of it, as described above, the staff determined that the application is consistent with guidance provided in NUREG-1536, Section 4.4.4 (Analytical Methods, Models, and Calculations) and therefore, meets the requirements of 10 CFR 236(f).

4.3 Thermal Evaluation for Short-Term Operations 4.3.1 Vacuum Drying The applicants methodology for performing vacuum drying is summarized in Section 4.11.1.3 of the FSAR. The applicant performed vacuum drying calculations for three heat load patterns (X, Y, and Z). The maximum temperatures of fuel and basket at the end of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of vacuum drying for these heat load patterns are reported in FSAR Table 4.11-4. As shown in this table, heat load pattern Z results in the highest fuel temperature at the end of vacuum drying but predicted temperature remains below the allowable limit described in the FSAR. Predicted results during cyclic vacuum drying show that temperature variations exceed the 117°F (65°C) threshold recommended in NUREG-1536. Section 8.6 of Chapter 8 Materials Evaluation provides the reasons why the staff has determined that exceeding the temperature variation threshold is acceptable in this instance.

The staff reviewed the applicants thermal evaluation of the MAGNASTOR storage system during vacuum drying. Based on the information provided in the application regarding the thermal analysis and evaluation, as described above, the staff determined that the application is consistent with guidance provided in NUREG-1536, Section 4.4.4 (Analytical Methods, Models, and Calculations) and therefore, meets the requirements of 10 CFR 236(f).

4.3.2 Onsite Transfer The applicants methodology for performing onsite transfer is summarized in Section 4.11.1.3 of the FSAR. The applicants calculated results are provided in FSAR Table 4.11-7. As shown in this table, predicted temperature remains below the allowable limit described in the FSAR for the administrative time use during transfer of the TSC to complete the operation.

The staff reviewed the applicants thermal evaluation of the MAGNASTOR storage system during on-site transfer. Based on the information provided in the application regarding the thermal analysis model, evaluation, and temperatures, the staff determined that the application is consistent with guidance provided in NUREG-1536, Section 4.4.4 (Analytical Methods, Models, and Calculations) and therefore, meets the requirements of 10 CFR 236(f).

4.4 Off-Normal and Accident Events 4.4.1 Off-Normal Events The applicant evaluated the following off-normal storage events: severe ambient temperature and half inlets blocked conditions. The off-normal event for variation in the ambient temperature only requires a change to the boundary condition temperature. For the half-blocked air inlets condition, the air inlet condition is modified to permit air flow through half of the inlet area. The applicant used heat load pattern Z to perform these analyses since it is the bounding heat load pattern. The temperatures of different components for off-normal storage conditions are provided in SAR Section 4.11.3.1. All component temperatures remain below the allowable limits described in the FSAR for off-normal conditions.

4.4.2 Accident Events The applicant evaluated the following storage accident events: maximum anticipated temperature, fire, and full blockage of air inlet vents. The applicant performed steady state analysis using the thermal model described in SAR Section 4.11.1 but modified accordingly to reflect the different event conditions. The temperatures of different components for accident conditions during storage are provided in SAR Section 4.11.4. All component temperatures remain below the allowable limits described in the SAR for accident conditions.

The staff reviewed the applicants thermal evaluation during off-normal and accident events.

Based on the information provided in the application regarding the thermal evaluation, the staff determined that the application is consistent with guidance provided in NUREG-1536, Section 4.4.4 (Analytical Methods, Models, and Calculations) and therefore, meets the requirements of 10 CFR 236(f).

4.5 Confirmatory Analyses The staff reviewed the applicants thermal models used in the analyses, checking the code input in the calculation packages submitted and confirming that the proper material properties and boundary conditions were used. The staff verified that the applicants selected code models and assumptions were adequate for the flow and heat transfer characteristics prevailing in the MAGNASTOR geometry for the analyzed conditions.

Engineering drawings were also consulted to verify that system geometry and dimensions were adequately translated to the thermal analysis models. The material properties presented in the FSAR were reviewed to verify that they were appropriately referenced and applied. In addition, the staff performed appropriate sensitivity analysis calculations to verify that applicants predicted results provide bounding predictions for all conditions analyzed in the application.

4.6 Evaluation Findings

F4.1 Chapter 2 of the FSAR describes SSCs important to safety to enable an evaluation of their thermal effectiveness. Cask SSCs important to safety remain within their operating temperature ranges.

F4.2 The MAGNASTOR storage system is designed with a heat-removal capability having verifiability and reliability consistent with its importance to safety. The cask system (TSC, transfer cask and Concrete overpack) is designed to provide adequate heat removal capacity without active cooling systems.

F4.3 The spent fuel cladding is protected against degradation leading to gross ruptures under long-term storage by maintaining cladding temperatures below 752°F (400°C).

Protection of the cladding against degradation is expected to allow ready retrieval of spent fuel for future processing or disposal.

F4.4 The spent fuel cladding is protected against degradation leading to gross ruptures under off-normal and accident conditions by maintaining cladding temperatures below 1058°F (570°C). Protection of the cladding against degradation is expected to allow ready retrieval of spent fuel for future processing or disposal.

F4.5 The staff finds that the thermal design of the MAGNASTOR storage system complies with the design requirements in 10 CFR 72.236 and that the applicable design and acceptance criteria have been satisfied. The evaluation of the thermal design provides reasonable assurance that the cask will allow for safe storage of spent nuclear fuel. This finding is reached based on a review that considered the regulation itself, appropriate regulatory guidance, applicable codes and standards, and accepted engineering practices.

Chapter 5 CONFINEMENT EVALUATION In MAGNASTOR Amendment No. 9, the changes proposed in the SAR did not change the confinement design/function of the TSC, which is tested to leaktight criteria, in accordance with ANSI N14.5, American National Standard for Leakage Tests on Packages for Shipment of Radioactive Materials,.

Since the design/function of the confinement system of the TSC did not accrue any changes, the NRC staff finds that the safety and regulatory compliance conclusions remain unchanged, thus the TSC remains in compliance with 10 CFR Part 72 and that the applicable design and acceptance criteria continue to be satisfied.

Chapter 6 SHIELDING EVALUATION The objective of this evaluation is to determine whether the shielding design of the MAGNASTOR dry cask spent fuel storage system for Amendment No. 9 meets the regulatory requirements. The review seeks to ensure that the shielding design is reasonably capable of meeting the dose requirements in 10 CFR 72.236(d).

The staff reviewed the application for Amendment No. 9. The staffs review includes the radiation source term determination and the radiation shielding design for all credible normal conditions and off-normal and accident events encountered during loading, handling, on-site transfer, storage, and retrieval analyses. This review also includes verification of computer modeling of the cask system for shielding analyses of the stainless steel MTC2 with a shielded closure; whether the MAGNASTOR dry cask spent fuel storage system with the new CC6 and contents meets the radiation protection requirements set forth in 10 CFR Part 72 and 10 CFR Part 20; and whether the design and operation of the MAGNASTOR storage system follow the ALARA principle.

6.1

System Description

6.1.1 Shielding Design Description The MAGNASTOR is a spent fuel dry storage system designed by NAC and is previously certified by the NRC under Docket No. 72-1031. The applicant submitted an application for an amendment to the CoC with new loading patterns for the B&W 15x15 fuels and a revised shielding design for the concrete overpack. Specifically, the proposed changes to the system design include:

1. The Table at the bottom of FSAR Figure 2.2-1 was replaced with a version that now includes all the zone loading patterns and maximum heat loads per storage location, including the four new loading patterns. Note, the NRC has previously approved patterns A, B, C, and D. Patterns X, Y, Z, and Z-Prime are the new loading patterns for B&W 15x15 fuel.

2. Added a new hybrid B&W 15x15 fuel assembly (BW15H5). Also, a new maximum enrichment level for the hybrid fuel assembly BW15H2 at 5 wt%

3. A short, augmented shielding segmented cask, CC6, with a 3-inch liner thickness and a revised lid design for the storage of high heat load, short cool time, B&W15x15 fuel assemblies and associated nonfuel hardware.

This revision request does not involve changes to the shielding structure for MTC2 and the 37 PWR basket. However, the applicant added three preferential heat loads for B&W 15x15 fuel assemblies to be placed into the 37 PWR fuel basket.

6.1.2 Design Features MAGNASTOR is a dry storage system consisting of a concrete cask, and a welded stainless-steel canister that has a welded closure to store up to 37 undamaged PWR spent fuel assemblies in the 37 PWR basket assembly. The system is also designed to store up to four damaged fuel assemblies in DFCs in the PWR damaged fuel basket assembly. The system also includes a transfer cask as a shielded lifting device designed to hold the canister during loading, transfer, and unloading operations. The fuels will be stored in four different preferential heat load configurations.

The TSC shell and the internal basket structure provides additional radial shielding. The transfer cask bottom shielding is provided by the TSC bottom plate after placement in the MTC2. The TSC closure lid provides radiation shielding at the top of the TSC.

6.2 Source Specification The applicant used the B&W15x15 PWR fuel specified in Revision 0 of the FSAR but with a minimum cooling time of 1.75 years and a maximum enrichment of 5.0 wt% for use in the SAS2H sequence of SCALE 4.4 to calculate the source terms.

As part of changes in Amendment No. 9, the applicant reduced the minimum cooling time of fuel and certain non-fuel hardware. This amendment includes a minimum cooling time of 1.75 years for B&W15x15 fuel in the 37 PWR basket for preferential loading patterns. For this amendment, the maximum exposure and minimum cooling times for non-fuel hardware for bounding conditions are given in Table 1-1 of Calculation Package No, 30076-5005 Revision 0. This reduction of minimum cooling time of fuel is consistent with the Three Mile Island operating history.

Each assembly is based on the maximum fuel and hardware masses and represents a conservative bounding value of fuel and hardware mass for that group. Table 5.2.3-1 and Table 5.2.3-2 of the SAR provide the essential characteristics of the fuel assemblies to be loaded in the MAGNASTOR system.

The applicant used the SAS2H sequence of the SCALE 4.4 computer code system to evaluate the source terms of each spent fuel assembly group. The 44GROUPNDF5 cross section library, which is composed primarily of ENDF/B-V cross-sections with limited ENDF/B-VI data for a limited number of isotopes, is employed to improve the calculation accuracy. The minimum cool time tables account for potential uncertainties in the source generation abilities of SAS2H at burnups greater than 45 gigawatt days per metric ton (GWd/MTU) by reducing allowed heat loads by 5%. Based on the code validation discussion in Section 5.2 of the MAGNASTOR FSAR Revision 8, a 5% uncertainty is applied to the heat loads for fuel burnups above 45 GWd/MTU.

6.2.1 Non-fuel Hardware components Neutron Sources Non-fuel hardware is allowed to be stored in the MPC-37 per TS Appendix B Tables B2-5 and B2-7 in Section 2.0. The primary neutron source strengths and neutron source assembly (NSA) strengths are 1x108 to 1x109 neutron/second when used initially in the reactor. These sources are depleted by decay or neutron absorption during its time in the reactors. Typical neutron sources are plutonium-Beryllium (Pu-Be), Americium-Beryllium (Am-Be), and Californium (Cf-252). Cf-252 has a half-life of 2.6 years and is significantly reduced during the period of reactor operation, however, Pu -239 and Am-241, which have a half-life of 87 and 433 years, respectively, do not diminish during reactor operations. A secondary neutron source is caused by the thermal activation of antimony (Sb) in the reactor. The Sb-Be source produces neutrons via absorption of gamma radiation by Sb-124 resulting in the emission of a neutron from the Be via decay. With a half-life of 2.7 years, this source diminishes after the source is removed from the reactor. There is only one neutron source allowed per fuel assembly and one per basket, and they must be located in one of the center nine fuel assemblies in the interior locations of the fuel basket in the canister. These sources provide less than 3% of the total contribution of neutron sources from fuel assemblies. Since assemblies at the outer ring of the fuel basket provide shielding, to the neutron sources in the inner rings, these neutron sources dont have any significant contribution to dose rates outside the canister.

Appendix B of the TS states that fuel assemblies containing burnable poison rod assemblies (BPRAs), thimble plug devices (TPDs) may be stored in any fuel assembly in any storage location. Fuel assemblies could contain axial power shaping rods (APSRs), rod Fuel cluster control assemblies (RCCAs), control element assemblies (CEAs), control rod assemblies (CRAs) (including, but not limited to those with hafnium), and only one Neutron Source Assembly (NSA) per cask. The contribution to the neutron sources in the fuel region from a single NSA is less than 3% of the fuel assembly contribution.

The applicant added the heat load from all non-fuel hardware to X, Y, Z, and Z-Prime patterns.

6.3 Shielding Model The applicant performed dose rate calculations using MCNP 6, with the ENDF/B-V cross-section sets code. The applicant includes statistical uncertainty in the reported dose rates. The statistical uncertainty of the calculated dose rate is part of the computational results of the MCNP code that employs the Monte Carlo method for solving neutron and gamma shielding problems.

The applicant imported the MCNP model for CC1 from Revision 0 of the FSAR. The model was modified to include the dimensions of CC6, new heat pattern, and modified source term for 5 wt% enrichment and a 1.75 year minimum cooling time. The model was changed to include the preferential loading zones described above, and canister port dimensions from drawings.

Table 4.1 of the Calculation Package No. 30076-5002, Rev. 2, MAGNASTOR TMI [Three Mile Island] Concrete Cask Preferential Pattern Shielding Analysis, shows updated parameters for this modeling.

The NRC staff reviewed the shielding model presented by the applicant and finds the reported shielding model for undamaged and damaged fuel is consistent with the previously approved models and, therefore, to be acceptable.

6.3.1 Response Methodology The comparison of response function (Section 5.8.2 of the SAR as approved in Amendment No.

0) which are summaries of dose calculations at each energy and MCNP calculations based on the complete gamma, neutron, and hardware gamma source spectra are shown in the figures of Section 5.12 of the SAR for the transfer cask and concrete cask systems. As evaluated and documented in the safety evaluation report for Amendment No. 0, the technical basis for this method is founded on the fact that the output of the MCNP code is normalized to per particle per energy and the geometric location of the sources in the cask. Once the dose rate from a particle at a specific energy or energy group and a segment of a fuel assembly and/or the non-fuel hardware (if applicable), is determined, the shielding calculations becomes simple mathematic manipulation. On these bases, the staff finds this method to be acceptable for this application.

6.3.2 Shielding Discussion and Dose Results The applicant calculated the dose rates for CC6, MTC2 transporting the TSC that contains 37 B&W 15x15 PWR fuel assemblies and updated non-fuel hardware content. The TSC is placed inside the MTC2 and sealed before being moved to a CC6 to be placed on the ISFSI pad. The CC6 currently has a 3 inch liner versus the previously approved 1.75 inch liner in CC1.

The applicant performed the dose calculations for the bounding preferential heat load of 52.8 kW (only 35.5 kW is allowed to be loaded in each cask). The minimum cooling time of 1.75 years, maximum burnup evaluated is 62 GWd/MTU, and a minimum initial enrichment of 3.5 wt% U-235.

The applicant calculated the dose rates for casks containing non-fuel assemblies at shorter cooling times, which is consistent with short cooling times of fuel assemblies.

The NRC staff reviewed the evaluations presented by the applicant and finds the reported dose rate calculations and the MCNP shielding analysis model for undamaged fuel, damaged fuel, and non-fuel hardware to be acceptable for the reasons discussed below.

6.4 Model Specification The staff reviewed the MCNP model of CC6 for four preferential loading patterns as depicted by MCNP VisEd in SAR Figures 5.12.2 and Calculation Package No. 30076-5002, Revision 2, and determined that the applicant modeled it in sufficient detail to accurately represent the CC6 concrete cask. The applicant states in Calculation Package No. 30076-5002, Revision 2 that it modeled CC6 the same as previously approved amendments with the only difference in this instance being the liner thickness, which increased to 3 inches from the previously approved thickness of 1.75 inches. The material composition and dimensions for both TSC and MTC2 are the same as previously approved amendments. The change in geometry and materials is limited to the addition of the axial power shaping rods (ASPRs) in the modeling of the CC6, TSC, and MTC.

In addition, the staff performed independent shielding calculations of the CC6, TSC, MTC2, using information from the SAR drawings and design information, as discussed in Section 5.2 of this SER and Section 5.12 of the SAR, and determined that the applicants model is appropriately representative and capable of calculating reasonably accurate dose rates for the 37 PWR system.

6.4.1 Preferential Loading Figure 5.12.4.1of the SAR depicted the four preferential loading patterns X, Y, Z and Z-Prime with a minimum cooling time of 1.75 years and maximum burnup of 62 GWd/MTU for eight zone patterns The spent fuel assemblies are sorted into groups according to the assembly types, i.e., PWR, and fuel and hardware masses to determine the bounding radiation source terms of the B&W 15x15 fuel assemblies to be loaded in the MAGNASTOR system. A hypothetical bounding fuel assembly was created for each assembly type, termed a hybrid fuel assembly. Each hybrid fuel assembly is based on the maximum fuel and hardware masses and presents a conservative bounding value of fuel and hardware mass of that group. Table 5.2.3-1 and Table 5.2.3-2 (MAGNASTOR FSAR Revision 5, ADAMS Accession No. ML17132A265) provide the essential characteristics of the fuel assemblies to be loaded in the MAGNASTOR system, as approved by NRC staff in the previous amendments.

The heat load for each pattern is depicted in Section 5.12.4 of the SAR. The maximum values are shown in this section for each assembly in each pattern and total heat load for each pattern, though heat loads of higher than 35.5 kW is not allowed in these patterns.

6.4.2 MAGNASTOR Transfer Cask Number 2 and Concrete Cask Number 6 Dose Rates The staff reviewed the updated dose rate tables in SAR Sections 5.12.4. The applicant represented the dose rates for fuel assemblies only in Table 5.12.4-1. In the Calculation Package No. 30076-5002, Revision 2, the applicant evaluated all other patterns in response to an NRC staff request for additional information. The results show (Section 6.2 of the report) that pattern X (pattern E in the TS) has higher dose rates, and it bounds all other patterns. Dose rates for MTC2 are higher for the side and top of the cask than the previous dose rates for MTC2, as shown in Tables 5.1.3-4 and 5.1.3-5. The top dose rate shows a 25% higher dose rate than the previous dose rate for MTC2. However, the dose rates are evaluated at a much higher heat load for preferential loading than dose rates in Tables 5.1.3-5 by approximately 48%

(52.5 kW versus 35.5 kW). Since a maximum heat load of 35.5 kW is allowed to be loaded, the applicant concluded the occupational dose will not increase, and the 25% higher dose rate is negated by a 48% decrease in allowed heat load.

The dose rates for CC6 are higher on the side and outlets of the cask for the bounding loading pattern. The cask side dose rate is the dominant contributor to the site boundary dose. The radial dose rate on the surface of CCC6 for preferential loading has an average dose rate of 56.0 mrem/hr, which is 3% higher than Table 5.8.3-5 in the Revision 8 of the FSAR (ADAMS Accession No. ML17038A505) for a 40 kW heat load that bounds dose for the site boundary.

The 56.0 mrem/hr dose rate is based on the heat load of 52.5 kW for bounding pattern X (pattern E in the TS). Since general licensees are only allowed to load a maximum heat load of 35.5 kW in the 37 PWR basket, which is less by a factor of 1.2 than 40 kW (40 kW/35.5 kW), the applicant concluded that site boundary dose rate based on the 40 kW from Revision 8 of the FSAR is still bounding 6.4.3 Non-fuel Hardware Burnable Poison Rod Assemblies The cask loaded with BPRA has been reevaluated using the same method as used in the previous Amendment No. 0 in Section 5.8.5 of the FSAR. The full cask load of 37 BPRAs for a minimum cooling time of 1.75 years and maximum exposure of 32.5 GWd/MTU was analyzed.

The results are provided in the SAR in Table 5.12.5.1 for the MTC2 and the CC6. The top dose rate increased by 11% for CC6 and 17% for the MTC2. The CC6 outlet dose rate is also increased by 11%, as shown in Table 5.12.5-1. The increase in the CC6 radial dose rate was negligible.

6.4.4 Nonfuel Hardware Reactor Control Elements The reactor control elements (CEAs and RCCAs) were reevaluated using the same method as Section 5.8.5 of the Revision 0 of the FSAR. The cooling times for this nonfuel hardware was reduced from the previous evaluation. The cooling times used in the re-evaluation are 3.75 years for 315 GWd/MTU and 1.75 years for 75 GWd/MTU, in comparison to 12 to 20 years in the previous evaluation. The 75 GWd/MTU with a cooling time of 1.75 years has a higher dose rate and is therefore bounding. The maximum dose rates increase by 13% for the MTC2 and 7% for CC6 for the sides when loaded with reactor control elements. The impact on dose rates is higher than evaluations in previous amendments since loading 37 CEAs in all fuel assembles verses the previously allowed nine positions in the center cells.

6.4.5 Axial Power Shaping Rods The black APSRs were initially used in the reactor, but later, it was replaced by gray APSRs.

(Black and grey APSR have different poisons and active lengths.) The gray APSRs have a larger exposure and shorter cooling time. ASPRs may fully withdraw during operation or be inserted at various phases of reactor operation. Exposure for APSRs is a sum of fuel assembly burnup during insertion plus 20% of burnup when it is fully withdrawn. The applicant used cobalt-60 with an impurity 0.8 grams of cobalt per kg of steel (g/kg) for generating the source term. The length of the gray APSRs is 63 inches. The applicant analyzed the full 37 APSRs.

Table 5.12.6-2 of the SAR shows the dose rates for the MTC2 and CC6. There is a 10%

increase in dose rates for the bottom of the MTC2 and CC6 inlet. The dose rate for the MTC2 in the radial direction is also increased by 21%.

6.4.6 Damaged Fuel Table 5.1.3-9 and Table 5.1.3-10 in Revision 5 of the FSAR (ADAMS Accession No. ML17132A265) summarized the maximum dose rates for the MTC and concrete cask for damaged fuel from previously approved amendments. The tables show that the maximum dose rate for damaged fuel is 10% less than for undamaged fuel for both the MTC and concrete cask.

The applicant, in the previous amendments, used the reconfigured source in damaged fuel locations, and the peak source was applied for all damaged fuel mass without considering shielding of the lower nozzle source by damaged fuel material at these damaged fuel locations.

The loading pattern X (pattern E in the TS) has 30% more heat load than the maximum of 35.5 kW allowed in CC6 by the CoC. Therefore, there is a significant margin when the system is evaluated with undamaged fuel only. The applicant concluded that loading damaged fuel doesnt produce a dose rate in excess of dose rates from undamaged fuel, because damaged fuel is only stored in four corners of the PWR-DF basket, and As discussed below, the increase in dose rate due to damaged fuel is compensated by the conservative nature of the source term for the pattern X, no additional evaluations beyond what was in the previous amendments is necessary.

The heat load of the bounding preferential loading pattern X is 30% more than the maximum allowed heat load of 35.5 kW in the 37 PWR basket. The delta of 30% is much higher than the increase in dose rate due to damaged fuel. Therefore, loading the damaged fuel doesnt produce a dose rate higher than all undamaged fuel in the preferential loading patterns.

6.5 Combined Dose rates for fuel assemblies and all nonfuel hardware The combined dose rates for the MTC2 and the CC6 cask are summarized in a table on Page 5.12.9-1 of SAR. The table shows the MTC2 radial, top and bottom average dose rates, and provides the bounding nonfuel component that contributes the most to the dose rate. For example, for transfer cask radial dose rate, the APSR is bounding for the nonfuel component that contributes more to the dose rate.

6.6 Impact on Occupational and Site Dose The applicant calculated the annual dose for a real individual at the control area boundary of a hypothetical ISFSI containing a 2x10 array of MAGNASTOR casks loaded with fuel assemblies having bounding source terms in terms of burnup, enrichment, and cooling time combinations and bounding exposure and cooling times for non-fuel hardware. The site boundary dose is dominated by dose from the side of the concrete cask. The site boundary dose is based on a 40 kW, uniform heat load as discussed in Section 5.8.3.5 in Revision 5 of the FSAR. From Table 5.12.9 in the SAR, for a 52.5 kW heat load source term of the loading pattern X, the concrete cask side average dose rate is 62.5 mrem/hr. This includes the assumption that 37 APSRS are loaded and each have a short cooling time. As stated in Section 5.8.3.5 of Revision 5 of the FSAR, the site boundary dose was calculated based on a total side dose rate of 60.8 mrem/hr for the 40 kW heat load. While the dose rate is a little lower than 62.5mrem/hr for a 52.5 kW heat load utilizing the bounding loading pattern X, the TS limits the total heat load in a cask to 35.5 kW. The heat load limit for loading all preferential patterns is 35.5 kW, and given this lower heat load limit, the previous evaluation for a 40kW heat load is still bounding. The MTC2 dose rates for pattern X and non-fuel hardware are higher than previous amendments.

Based upon this analysis, the applicant has demonstrated, and the staff concurs that the MAGNASTOR dry cask storage system meets the radiation protection requirements of 10 CFR 72.236(d).

6.7 Confirmatory Review and Analysis The staff reviewed the applicants shielding analysis and found it acceptable. The maximum dose rates meet the limits defined by 10 CFR 72.236(d). The staff reviewed the radiation shielding evaluations, including the calculations of the sources and the dose rates for the transfer cask and the concrete overpack as well as the annual dose at the controlled area boundary. The staff independently calculated source terms for the bounding PWR 15x15 fuel assemblies using combinations of different enrichments, burnups, and cooling times. The staff performed confirmatory analyses of the dose rates for the MTC2. The staff also performed confirmatory analyses of the dose rates for CC6. Using irradiation parameter assumptions similar to the applicants, the staff obtained bounding source terms that were similar to or bounded by those determined by the NAC and therefore finds the applicants result acceptable.

The staff finds that the applicant has correctly assessed the bounding dose rates for all proposed contents, as defined in Tables 5.12.4.1, 5.12.5.1, 5.12.6.1.5.6.12.7.1 and 5.12.7.2 of their SAR to be acceptable. Based on this review and analyses, the staff concludes that the applicant has demonstrated that the MAGNASTOR dry cask storage system meets the radiation protection requirements of 10 CFR 72.236(d).

6.8 Evaluation Findings

The staff reviewed the applicants shielding analyses for the amended MAGNASTOR dry storage system design and finds that the approaches and methodologies used in these calculations and the results are acceptable for the MTC and CC6 system design. Based on its review of the information and representation provided by the applicant, the staff concludes that the requested change meets the regulatory limits and the acceptance criteria specified in NUREG-1536 and provides reasonable assurance for the safe transfer and storage of the spent fuel and non-fuel hardware as specified in the Technical Specifications for the system. On these bases, the staff finds:

F6.1 Chapter 5 of the MAGNASTOR SAR sufficiently describes the shielding design bases and design criteria for the structures, systems, and components important to safety.

F6.2 The MAGNASTOR system radiation shielding and transfer cask (MTC) and confinement features are sufficient to meet the radiation protection requirements of 10 CFR Part 20, and 10 CFR 72.236(d).

F6.3 The staff concludes that the shielding and radiation protection design features of the MAGNASTOR system, including the concrete cask, the transfer cask, and the TSC, comply with 10 CFR Part 72, and that the applicable design and acceptance criteria have been satisfied. The evaluation of the shielding and radiation protection design features provides reasonable assurance that the system will provide safe transfer and storage of spent fuels. This finding is based on a review that considered the regulation itself, the appropriate regulatory guides, applicable codes and standards, the applicants analyses, the staffs confirmatory analyses, and acceptable engineering practices.

Chapter 7 CRITICALITY EVALUATION Staff reviewed the amendment request to determine if the MAGNASTOR system continues to maintain its authorized contents in a subcritical configuration under all credible normal, off-normal, and accident events encountered during the handling, loading, transfer, and storage of spent nuclear fuel. Staff reviewed the applicants criticality safety analysis to ensure that all credible bounding scenarios were adequately identified and any potential consequences on the criticality safety of the MAGNASTOR dry cask storage system continues to meet the regulatory requirements of 10 CFR 72.124 and 72.236. The conclusions of the staff are based on the information provided by the applicant and the supporting calculations for the addition of: (1) a new hybrid B&W 15x15 fuel assembly (BW15H5); and (2) a new maximum enrichment of 5.0 wt% 235U for the hybrid BW15H2 fuel assembly. The hybrid BW15H2 assembly requires a new minimum soluble boron concentration of 2650 ppm 10B and neutron absorber panels that that have a minimum areal density of 0.036 10B g/cm3.

7.1 Criticality Design Criteria and Features The MAGNASTOR storage system consists of a storage canister, a concrete storage case, and a lead-shielded transfer cask. Criticality safety of the cask is provided by a combination of fissile mass and enrichment controls, geometry control, and fixed neutron absorbers in the basket. The MAGNASTOR system may also contain DFCs to store damaged PWR fuel.

Fixed neutron absorber sheets are attached to the walls of the fuel assembly tubes and sit between each fuel assembly in the basket. PWR fuel requires the use of soluble boron in the water that is used to flood the canister during loading and unloading operations. The minimum soluble boron content is based on the assembly type and the maximum initial assembly enrichment.

Since the previously approved cask design is not altered by this amendment, staff evaluated the addition of the BW15H5 assemblies, the increased enrichment of the BW15H2 assemblies, and the increased soluble boron level to address the use of the TSC for damaged BW15H2 assemblies. The applicant made changes to the CoC and the TS to allow these additions to the MAGNASTOR storage system.

Staff reviewed the model descriptions provided by the applicant and their assumptions and finds that they are consistent with the description of the design shown in the drawings and the contents provided in the SAR. Staff also evaluated the information the applicant provided in the SAR and found the criticality calculations were sufficiently detailed to support the staff evaluation. Based on this review, the staff finds that the applicant adequately met the requirements of 10 CFR 72.236.

7.2 Fuel Specifications Consistent with previous amendments, the applicant identified the proposed new fuel contents based on the specified fuel type and identified the conservative bounding values for the criticality significant parameters for each fuel. The hybrid BW15H2 fuel assembly parameters are unchanged with the exception of an increase in enrichment to 5.0 wt% 235U. The new hybrid BW15H5 fuel assemblies have the same number of fuel rods and guide tubes as other B&W 15x15 fuel assemblies. However, they are slightly different than those previously approved and have a minimum clad outer diameter (OD) of 0.422 inches, a minimum clad thickness of 0.0243 inches, a maximum pellet OD of 0.3659 inches, and a maximum enrichment of 5.0 wt%

235U. Staff reviewed the FSAR and the proposed fuel specifications in the TS and found the applicant adequately specified the proposed fuel specifications that could impact the criticality safety of the MAGNASTOR storage system.

7.3 Model Specifications The applicant evaluated the storage of both damaged and undamaged fuel assemblies in Amendment No. 3 and Amendment No. 5 (ADAMS Package Accession Nos. ML13207A245 and ML15180A364, respectively) to the MAGNASTOR system, and these are unchanged in this amendment. The applicant used the same criteria to evaluate the new models to support the additional fuel types specified.

7.4 Criticality Analysis The applicant requested the addition of a new hybrid fuel type, the BW15H5 and changes to the maximum enrichment of the BW15H2 under this amendment. The applicant performed a criticality evaluation that demonstrated the proposed new fuel contents for both damaged and undamaged fuel continue to remain under the established upper subcritical limit (USL) of 0.9372. The applicant evaluated the fuel under the same general conditions and assumptions used in previous amendments except for an increased neutron absorber areal density of 0.036 10B g/cm3 and an increased minimum soluble boron concentration of 2650 ppm 10B.

These changes allow for the storage of assemblies using the increased enrichment of the hybrid BW15H2 assemblies and the new hybrid BW15H5 assemblies, as well as damaged hybrid BW15H2 assemblies while in the TSC, respectively, to maintain the maximum keff of each assembly type to remain below the USL. The applicant placed the new hybrid undamaged assemblies in a DFC, evaluated the effect of loose rods without cladding (damaged fuel), and modeled loose rods as a homogenized mixture of fuel and moderator at various heights within the DFC cavity (rubble). Staff determined this evaluation to be acceptable due to the conservative assumptions used by the applicant, such as the higher enrichment and assuming loose rods with no cladding, as well as a mixture of fuel and water within the DFC cavity. In all instances the applicant calculated keff to be below the USL, and staff found this acceptable.

The applicant evaluated the new and revised hybrid fuel assemblies for both the undamaged and damaged configurations. For the undamaged configuration, hybrid BW15H2 and BW15H5 assemblies were modeled both with and without inserts at an enrichment of 5.0 wt% 235U. The applicant calculated the most reactive configuration for the BW15H2 assemblies at 2600 ppm 10B to have keff + 2 = 0.93400, and for the BW15H5 assemblies at 2500 ppm 10B the keff + 2 = 0.93113. For the damaged configurations, the applicant evaluated the three configurations listed in the above paragraph and determined that the maximum keff + 2 = 0.93645 for the BW15H2 fuel at an enrichment of 5.0 wt% 235U, a soluble boron concentration of 2650 ppm 10B, and modeled the fuel as a mixture of fuel and water at a mix height of 70%. The applicant calculated a maximum keff + 2 = 0.93265 for BW15H5 fuel at an enrichment of 4.9 wt% 235U, a soluble boron concentration of 2500 ppm 10B, and modeled the fuel as a mixture of fuel and water at a mix height of 80%. The applicant evaluated a wide range of mixture configurations as shown in Table 6-2 and Table 6-3 of the supporting calculations to the SAR (Calculation Package Number 30076-6001, Criticality Related TMI Site Specific Fuel Parameter Update) to find the maximum reactivity of damaged fuel contents in the DFC for both new fuel types. In all cases, the applicant demonstrated the reactivity was below the USL.

The applicant used the MCNP5, a three-dimensional Monte Carlo code, with continuous neutron energy cross-sections. This code was developed by the Los Alamos National Laboratory for performing criticality analyses and was used in calculations in previous amendments to the MAGNASTOR cask system. Staff finds its use acceptable for this system since the MCNP code and cross-section libraries are benchmarked by comparison to a wide range of critical experiments for light-water reactor fuel in storage and transportation packages. The applicant also performed calculations showing that the MAGNASTOR system will continue to meet the design criterion of keff + 2 < USL when loaded with the authorized contents as specified in the SAR and proposed TSs.

7.5 Criticality Evaluation Summary All of the applicants models for the new hybrid fuel types are based on the engineering drawings in the SAR and previously reviewed models submitted as part of supplemental calculations. The design-basis, off-normal, and accident events do not affect the design of the cask with regards to maintaining the MAGNASTOR cask system in a subcritical configuration.

This means the calculation models for the normal, off-normal, and accident conditions are the same. Staff imported sample input files provided by the applicant in support of its supplemental calculations to confirm the results provided by the applicant. For these reasons, staff finds that the applicants evaluation of the criticality design demonstrates that the MAGNASTOR cask storage system will continue to allow for the safe storage of spent fuel. This finding is reached on the basis of a review that considered the regulation itself, appropriate regulatory guides, applicable codes and standards, and accepted engineering practices.

7.6 Evaluations Findings Staff reviewed the information provided in Amendment No. 9 of the MAGNASTOR cask system application and determined that it is in compliance with the requirements in 10 CFR 72.124, and 10 CFR 72.236(c). Staff also determined that the results of the applicants evaluation of the new hybrid BW15H5 fuel assembly and the new maximum enrichment of 5.0 wt% 235U for the hybrid BW15H2 fuel assembly, as described in this application, remain less than the USL of 0.9372 for each of the evaluated cases. The applicant incorporated a number of conservative assumptions and evaluated the fuel over a range of bounding credible scenarios. Limits are imposed for each fuel in regard to the minimum 10B concentration in the absorber sheets, soluble boron concentrations in the pool, and allowable enrichments. As a result, staff has reasonable assurance that the MAGNASTOR spent fuel dry cask storage system containing either of the hybrid BW15H5 and BW15H2 fuels, as described in this amendment to the application, will remain safe while in storage. Specifically, the applicants nuclear criticality safety evaluation demonstrates that the MAGNASTOR spent fuel dry cask storage system will continue to meet the relevant regulatory requirements and the staff finds the following:

F7.1 The applicant described structures, systems, and components important to criticality safety in sufficient detail in the SAR to enable an evaluation of their effectiveness.

F7.2 The cask and its spent fuel transfer systems are designed to be subcritical under all credible conditions.

F7.3 The criticality design is based on favorable geometry, fixed neutron poisons, and soluble poisons of the spent fuel pool. An appraisal of the fixed neutron poisons has shown that they will remain effective for the term requested in the application and there is no credible way for the fixed neutron poisons to significantly degrade during the requested term in the application; therefore, there is no need to provide a positive means to verify their continued efficacy as required by 10 CFR 72.124(b).

F7.4 The applicants analysis and evaluation of the criticality design and performance have demonstrated that the cask will enable the storage of spent fuel for the term requested in the application.

Chapter 8 MATERIALS EVALUATION The staff evaluated the materials changes associated with the new CC6 version of the concrete cask, the new hybrid B&W 15x15 (BW15H5) fuel assembly contents, and minor changes related to these new SSCs.

The staff notes that the new SSCs introduced in the amendment are variations of SSCs that have been previously approved in the MAGNASTOR CoC. The materials, fabrication, and examination requirements of the CC6 and BW15H5 fuel assemblies are unchanged. The concrete cask is manufactured from reinforced concrete and carbon steel structures (e.g., liner, standoffs, baseplate). The new fuel assembly type comprises zirconium alloy cladding and nonfuel component materials that are largely constructed of stainless steel, carbon steel, Inconel, and zirconium-based alloys.

The staff reviewed the amendment changes to ensure that the applicant adequately described and evaluated materials performance, including whether the new cask and contents introduce any changes in the service environment (e.g., temperature) that could affect materials performance of all important-to-safety SSCs.

8.1 Drawings The applicant provided new drawings in Section 1.8 of the SAR to incorporate the CC6 version of the concrete cask, and it provided several updates to existing drawings to incorporate minor changes to other SSCs (e.g., fuel tube assemblies, basket assemblies, and damaged fuel cans). The drawings include a parts list that provides the material specification of each component, and they also provide the welding, examination, and coating requirements. The staff notes that the level of detail in the new and revised drawings are consistent with those of the previously approved drawings. 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, fabrication, and examination requirements, and, therefore, the staff finds them to be acceptable.

8.2 Codes and Standards The staff verified that the new CC6 uses the same American Concrete Institute construction codes, ASTM International steel materials, welding requirements, and examination methods as the previously approved cask versions. Similarly, minor changes to other components (e.g. dimensions) did not affect the applicable codes and standards. Therefore, the staff finds the materials codes and standards to be acceptable.

8.3 Mechanical Properties The applicant did not make any changes to the mechanical properties used in the structural analyses. However, the staff reviewed the applicants thermal analysis to ensure that those mechanical properties remain valid under the service conditions associated with the new CC6, BW15H5 fuel contents, and the fuel zoned loading patterns that were added in the amendment.

In SAR Section 4.11, the applicant evaluated the maximum temperatures of the fuel cladding, fuel basket, canister shell, and cask concrete under normal, off-normal, and accident conditions.

The staff reviewed the applicants analysis and verified that the component temperatures remain below each of the materials allowable service temperatures. Therefore, the staff finds the mechanical properties used in the applicants structural analysis to be acceptable.

8.4 Thermal Properties In SAR Section 8.3, the applicant added data on the thermal conductivity of the Type 2 neutron absorber that was introduced in the amendment. The absorber materials were unchanged (Boral, metal matrix composite, and borated aluminum have all been previously approved);

however, the Type 2 variation designates a higher thermal conductivity to meet the needs of the new BW15H5 fuel contents and loading patterns. The staff reviewed proprietary materials test data and verified that the applicant used appropriate thermal conductivity values in its thermal analysis for the Type 2 absorber; therefore, the staff finds the neutron absorber thermal properties to be acceptable.

In addition, in SAR Tables 8.3-30 and 8.3-33, the applicant added footnotes to the thermal emissivity values for nickel-plated steel fuel baskets and shield plates and the stainless steel canister shell, respectively. These footnotes clarified the values that are applicable to the thermal analysis of the new CC6 and its contents. The staff reviewed literature data and proprietary test reports and verified that the applicant identified appropriate emissivity values for use in its thermal analysis; therefore, the staff finds the thermal properties of the nickel-plated steel and stainless steel to be acceptable.

8.5 Corrosion Resistance and Content Reactions The staff reviewed the amendment changes and verified that they do not introduce any adverse corrosive or other reactions that were not previously considered in the staffs prior review of the MAGNASTOR CoC. The materials of construction and the service environments are bounded by those that were previously evaluated in the CoC. Therefore, the staff finds the applicants evaluation of corrosion resistance and potential adverse reactions to be acceptable.

8.6 Spent Fuel As described in SAR and Table B2-3 of the TS, the new BW15H5 fuel assembly is a subtype of B&W 15x15 assemblies previously approved in the MAGNASTOR CoC. The new subtype introduces a unique combination of rod pitch, cladding dimensions, and fuel pellet diameter.

The materials of construction for the fuel assemblies and cladding are unchanged in the amendment. In addition, as discussed above in the staffs evaluation of mechanical properties, the staff verified that the fuel cladding temperatures remain below those associated with the previously approved casks and contents, as well as below the allowable cladding temperatures recommended in NUREG-1536.

In SAR Section 4.11, the applicant revised the calculated fuel cladding temperatures during drying operations. The revisions include temperature changes that exceed the 65°C threshold recommended in NUREG-1536 to limit potential detrimental effects of hydride reorientation on cladding mechanical performance. However, the staff notes that the exceedance of the recommended thermal cycling threshold was previously reviewed and approved by the staff in CoC No. 1031, Amendment No. 0 and is defined in the existing MAGNASTOR TSs. In addition, the staff notes that more recent research that subjected zirconium cladding to multiple, more excessive, thermal cycles (between 100°C and 230°C) has found that such cycling does not significantly affect the extent of hydride reorientation and is not expected to detrimentally affect cladding performance under the bending loads associated with drop accidents (Billone, 2014; NRC, 2017). Therefore, based on the staffs prior approval of the applicants thermal cycling criteria and more recent research that provides additional support for the use of that criteria, the staff finds the revised cladding temperature changes during fuel drying operations to be acceptable.

8.7 Materials Findings The staff concludes that the changes associated with the amendment ensure adequate materials performance, considering materials properties, corrosive and other adverse reactions, and the effects of drying practices on fuel cladding integrity.

F8.1 The applicant has met the requirements in 10 CFR 72.236(b). The applicant described the materials design criteria for SSCs important to safety in sufficient detail to support a safety finding.

F8.2 The applicant has met the requirements in 10 CFR 72.236(g). The properties of the materials in the storage system design have been demonstrated to support the safe storage of spent fuel.

F8.3 The applicant has met the requirements in 10 CFR 72.236(h). The materials of the spent fuel storage container are compatible with their operating environment such that there are no adverse degradation or significant chemical or other reactions.

F8.4 The applicant has met the requirements in 10 CFR 72.236(a) and 10 CFR 72.236(m).

spent fuel specifications have been provided and adequate consideration has been given to compatibility with retrieval of stored fuel for ultimate disposal.

References Billone, M.C., T.A. Burtseva, Z. Han, and Y.Y. Liu. 2014. Effects of Multiple Drying Cycles on High-Burnup PWR Cladding Alloys, DOE Used Fuel Disposition Report, FCRD-UFD-2014-000052, ANL Report ANL-144/11.

NRC. 2017. Mechanical Fatigue Testing of High-Burnup Fuel for Transportation Applications, NUREG/CR-7198, Rev. 1, Washington, DC, October 2017. ADAMS Accession No. ML17292B057.

Chapter 9 OPERATING PROCEDURES EVALUATION There were no requested changes to the operating procedures and none of the changes in this amendment affect the operating procedures section.

Chapter 10 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM EVALUATION NAC revised the acceptance tests to include the minimum required thermal conductivity for Type 1 and Type 2 neutron absorbers, consistent with the evaluations in Chapters 4 and 8, above.

Chapter 11 RADIATION PROTECTION EVALUATION Since the average dose rates around the cask, when loaded with fuel at 35.5 kW are bounded by the dose rates in Revision 5 of the FSAR, the revisions requested by NAC do not affect the radiation protection components of the system and do not alter the staffs previous evaluation of radiation protection of the MAGNASTOR system. Therefore, the staff did not reevaluate this area for this revision request.

Chapter 12 ACCIDENT ANALYSES EVALUATION The staff reviewed the changes in Chapter 12 of the SAR and determined that all the changes are editorial in nature. Therefore, the revisions requested by NAC do not affect the accident analysis evaluation for the system and do not alter the staffs previous evaluation of the accident analyses for the MAGNASTOR system. The staff determined that reorganization of the Appendix B of the TS does not modify the TS and maintains the TS requirements for previously approved contents.

Chapter 13 TECHNICAL SPECIFICATIONS AND OPERATING CONTROLS AND LIMITS EVALUATION In addition to the changes to the TSs due to addition of the CC6, the new hybrid fuel assembly and its preferential loading patterns, NAC reorganized the Appendix B to the TSs to limit duplication of material and simplify Appendix B such that it is easier for licensees to use. The only changes made to Appendix A of the TS were revision of the amendment number and inclusion of registered trademark symbol on the word MAGNASTOR. NRC staff reviewed the revised TSs to ensure that previously approved TS were properly incorporated, and that none were omitted or revised.

F13.1 The staff concludes that the conditions for the MAGNASTOR storage system identify necessary technical specifications to satisfy 10 CFR Part 72 and that the applicable acceptance criteria have been satisfied.

Chapter 14 QUALITY ASSURANCE PROGRAM EVALUATION There were no requested changes to NACs quality assurance program and none of the changes requested affect the quality assurance program.

Issued with Certificate of Compliance No. 1031, Amendment No. 9, on ___draft______