Text

Final Safety Evaluation Report for the Standardized Nuhoms Horizontal Modular Storage System for Irradiated Nuclear Fuel
	ML23293A300
Person / Time
Site:	07201004
Issue date:	11/08/2023
From:	; Storage and Transportation Licensing Branch
To:	; TN Americas LLC
Shared Package
ML23293A298	List: ML23293A299; ML23293A300; ML23293A301; ML23293A302; ML23293A303; ML23293A304;
References
	EPID L-2022-LLA-0079, CAC 001028
	Download: ML23293A300 (1)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 FINAL SAFETY EVALUATION REPORT TN AMERICAS, LLC.

STANDARDIZED NUHOMS HORIZONTAL MODULAR STORAGE SYSTEM FOR IRRADIATED NUCLEAR FUEL DOCKET NO. 72-1004 AMENDMENT NO. 18

i TABLE OF CONTENTS

SUMMARY

................................................................................................................................ 4 1.0 GENERAL INFORMATION............................................................................................ 5 1.1 Description of 24PTH System Changes...................................................................... 5 1.1.1 24PTH Basket Design......................................................................................... 5 1.1.2 DSC Content Changes....................................................................................... 6 1.1.3 Horizontal Storage Module Changes................................................................ 6 1.1.4 Transfer Cask Changes..................................................................................... 6 1.2 System Drawings.......................................................................................................... 6 1.3 Technical Qualifications of Applicant......................................................................... 7 1.4 Changes to Standardized NUHOMS Certificate of Compliance............................... 7 1.5 Changes to Standardized NUHOMS Technical Specifications................................ 7 1.7 Evaluation Findings..................................................................................................... 7 2.0 DRY STORAGE FACILITY SITE CHARACTERISTICS EVALUATION......................... 7 3.0 PRINCIPAL DESIGN CRITERIA EVALUATION............................................................ 7 4.0 STRUCTURAL EVALUATION....................................................................................... 8 4.1 System Description...................................................................................................... 8 4.2 Evaluations of the 24PTH DSC Type 3 Basket........................................................... 8 4.2.1 Evaluation of ANSYS FE Model Analysis......................................................... 9 4.2.2 Evaluation of End Drop Analysis Under Accident Condition.........................12 4.2.3 Evaluation of Creep Model Analysis for Long Term Storage.........................12 4.2.4 Evaluation of Crack Propagation and Growth Analysis.................................12 4.3 Evaluation Findings....................................................................................................12 4.4 References...................................................................................................................13 5.0 THERMAL EVALUATION.............................................................................................13 5.1 Change No. 1 - Lighter 24PTH Type 3 Basket Design..............................................14 5.1.1 Storage Analysis of the 24PTH DSC with the Type 3 Basket.........................14 5.1.2 Transfer Analysis of the 24PTH DSC with the Type 3 Basket........................15 5.1.3 Thermal Model Audit.........................................................................................15 5.2 Change No. 2 - Delete Appendix A Requirement for Initial HSM Delta T Measurement with Loaded DSC.................................................................................15

ii 5.3 Change No. 7 - Clarify 24PTH-S-LC DSC Transfer Time Limit in TS Appendix B LCO 3.1.3, Consistent with the Existing UFSAR Analysis........................................16 5.4 Evaluation Findings....................................................................................................16 6.0 SHIELDING EVALUATION...........................................................................................17 6.1 Modified 24PTH Basket and 24PTH-S-LC DSC Shield Plug Lead............................18 6.1.1 Shielding Design Description...........................................................................18 6.1.2 Radiation Source Definition..............................................................................20 6.1.3 Shielding Model Specification..........................................................................21 6.1.4 Shielding Analyses...........................................................................................23 6.1.5 CoC Appendix A 24PTH DSC Dose Rate Limits..............................................26 6.2 HSM Concrete Cement Mixture Change....................................................................28 6.3 Evaluation Findings....................................................................................................28

7.0 CRITICALITY EVALUATION

........................................................................................29 7.1 Criticality Design Criteria and Features.....................................................................29 7.2 Proposed Change No. 1 - 24PTH Type 3 Basket Design..........................................30 7.2.1 Fuel Specification..............................................................................................30 7.2.2 Model Specification...........................................................................................30 7.2.3 Criticality Analysis............................................................................................31 7.3 Evaluation Findings....................................................................................................33 8.0 MATERIALS EVALUATION.........................................................................................34 8.1 Change No. 1 - Lighter 24PTH Type 3 Basket Design..............................................35 8.2 Change No. 4: NG-4231.1 Code Alternative Addition to Appendix C.......................37 8.3 Change No. 7: Clarify Appendix B TS LCO 3.1.3 Time Limit....................................37 8.4 Additional Scope Change No. 1: Change Regarding Blended Portland-Limestone Cement.........................................................................................................................37 8.5 Management of Aging Mechanisms and Effects.......................................................40 8.6 Findings.......................................................................................................................41 9.0 CONFINEMENT EVALUATION....................................................................................41 9.1 Confinement System Evaluation................................................................................42 9.2 Evaluation Findings....................................................................................................42 10.0 RADIATION PROTECTION EVALUATION..................................................................43

iii 10.1 Modified 24PTH Basket and 24PTH-S-LC DSC Shield Plug Lead............................44 10.2 Changes to CoC Appendix B Section 4.3.2...............................................................45 10.3 HSM Concrete Cement Mixture Change....................................................................46 10.4 Evaluation Findings....................................................................................................47 11.0 OPERATING PROCEDURES.......................................................................................47 11.1 Change No. 9: Short Term Operational Controls......................................................47 11.2 Evaluation Findings....................................................................................................49 12.0 CONDUCT OF OPERATIONS......................................................................................49 12.2 Evaluation Findings....................................................................................................51 13.0 QUALITY ASSURANCE EVALUATION.......................................................................51 14.0 ACCIDENT ANALYSIS EVALUATION.........................................................................51 15.0 CONDITIONS FOR CASK USE - TECHNICAL SPECIFICATIONS..............................51 15.1 Conditions for Use......................................................................................................51 15.2 Standardized NUHOMS Certificate of Compliance Changes..................................51 15.3 Technical Specifications Changes.............................................................................52 15.4 Evaluation Findings....................................................................................................53

16.0 CONCLUSION

..............................................................................................................54

4 PRELIMINARY SAFETY EVALUATION REPORT Docket No. 72-1004 Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel Certificate of Compliance No. 1004 Amendment No. 18

SUMMARY

This safety evaluation report (SER) documents the U.S. Nuclear Regulatory Commission (NRC) staffs review and evaluation of the request to amend Certificate of Compliance (CoC) No. 1004 for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel.

By application dated May 20, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML22140A025 and ML22213A161), as supplemented in letters dated August 12, 2022 (ML22224A041), January 20, 2023 (ML23020A920), January 27, 2023 (ML23027A056), February 16, 2023 (ML23047A028), March 8, 2023 (ML23191A075), and April 5, 2023 (ML23095A100), TN Americas LLC, from here on referred to as the applicant, requested that NRC amend the CoC to include the following changes:

Change No. 1:

Provide for a 24PTH improved basket design (Type 3) using staggered plates similar to EOS-37PTH to simplify construction, reduce weight, and improve fabricability.

Change No. 2:

Deletion of Appendix A inspections, tests, and evaluations (ITE) requirement for initial horizontal storage module (HSM) delta temperature (T) measurement with a loaded dry shielded canister (DSC).

Change No. 3:

Clarifies Appendix B technical specification (TS) Section 4.3.2 language related to transfer casks with liquid neutron shields regarding the OS197L transfer cask (TC),

which is significantly different than other TC models.

Change No. 4:

Updated Appendix C American Society of Mechanical Engineers (ASME) Code Alternatives Table C-12 to add code alternative NG-4231.1 Change No. 5:

Change Appendix B TS Section 4.3.2, first paragraph, by removing a reference to Part 20 of Title 10 of the Code of Federal Regulations (10 CFR Part 20) to clarify language indicating that the site specific evaluation in accordance with 10 CFR 72.212 is to demonstrate compliance with 10 CFR 72.104.

Change No. 6:

Clarify TC/DSC annulus draining language in Appendix B TS Section 4.3.2 within the last paragraph.

Change No. 7:

Clarify in Appendix B TS LCO 3.1.3 that there is no transfer time limit associated with the 24PTH-S-LC DSC, consistent with existing updated final safety analysis report (UFSAR) analysis.

5 Change No. 8:

Editorially correct certificate of compliance (CoC) name/address information by adding missing space between 7160 and Riverwood Drive. Note that staff made this editorial change to the CoC. No evaluation required.

Change No. 9:

Incorporate administrative controls during short duration independent spent fuel storage installation (ISFSI) handling operations that are unanalyzed for tornado hazards in accordance with the guidance contained in NRC EGM 22-001, Enforcement Discretion for Noncompliance of Tornado Hazards Protection Requirements at Independent Spent Fuel Storage Installations.

Additional Scope Change No. 1:

The applicant also requested, as an additional scope item, a change to the HSM concrete, to allow use of different cement, which is a blended Portland cement meeting the requirements of the American Society for Testing and Materials (ASTM) C595 standard.

The amended CoC, when codified through rulemaking, will be denoted as Amendment No. 18 to CoC No. 1004. This SER documents the staffs review and evaluation of the proposed amendment. The staff followed the guidance of NUREG-2215, Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities, when performing technical reviews of spent fuel storage and transportation packaging licensing actions.

The staff's evaluation is based on a review of the applicants application and whether it meets the applicable requirements of 10 CFR Part 72 for dry storage of spent nuclear fuel. The staffs evaluation focused only on modifications to the CoC, and TS requested in the amendment as supported by the submitted revised UFSAR (see ADAMS Package Accession Nos.

ML22213A161, ML22224A041, ML23020A920, ML23027A056, ML23047A028, ML23191A075, and ML23095A100) and did not reassess previous revisions of the UFSAR nor previous amendments to the CoC.

1.0 GENERAL INFORMATION 1.1 Description of 24PTH System Changes 1.1.1 24PTH Basket Design The applicant proposed adding a third basket design to the existing 24PTH system. The new basket design, which was designated as the Type 3 basket, utilized composite plate compartments. The applicant plans to incorporate a high Boron-10 (10B) content, i.e., Type D, into the new basket designs poison plates. The applicant only plans to incorporate the high 10B content into the Type 3 basket. These changes resulted in the Type 3D basket design which is the seventh basket design for the 24PTH system. The applicant based the Type 3D basket on the EOS-37PTH basket design approved in Amendment No. 1 to CoC No. 1042 (ADAMS Package Accession No. ML20136A048). The 24PTH Type 3 basket structure consists of composite plates of steel, neutron poison, and aluminum plates that fit together to form a grid structure. This differs from the Type 1 and Type 2 basket designs where each fuel tube is welded together at selected elevations along the axial length of the basket through stainless steel insert plates, which separate the aluminum and poison plates, arranged in an egg crate configuration. The 24PTH Type 3 basket uses extruded aluminum R45 transition rails with open sections. Internal steel angles reinforce the transition rails and provide structural strength.

6 1.1.2 DSC Content Changes The applicant proposed to increase the allowable weight limit for the contents (fuel assembly and control components) in the 24PTH DSC with a Type 3 basket from 1682 lbs. to 1715 lbs.

The applicant stated that this 33-pound increase in the maximum weight for the contents was the result of the weight reduction associated with the new Type 3 basket. In addition, the applicant explained in a supplemental submittal that the new maximum weight for the contents was derived using the most conservative fuel assembly weight with the most conservative control component weight as well as the uncertainties in those weights (ADAMS Package Accession No. ML23191A075). The staff evaluated this change and described their findings in SER sections 4, 5, 6 and 7.

To provide a consistent content weight description, the applicant proposed to delete the total weight of the Failed Fuel Can (FFC) plus all its contents from table 1-1l. For completeness, the applicant also proposed to delete the total weight requirement for the FFC plus all its contents from tables 1-1e and 1-1t for the 32PT DSC and the 61BTH DSC, respectively. The applicant asserted that this level of detail is consistent with weight requirements contained in the TS for CoC No. 1029 and CoC No. 1042. Staff confirmed that these changes provide a level of detail consistent with other staff-approved TS.

1.1.3 Horizontal Storage Module Changes The applicant included a change in this amendment to the HSM concrete, to allow use of different cement, which is a blended Portland cement meeting the requirements of the American Society for Testing and Materials (ASTM) C595 standard. The applicant updated a table in FSAR section P.4.1 to identify the HSM designs in which a 24PTH DSC utilizing the new Type 3 basket is authorized for storage. The applicant also added text clarifying that the previously authorized Type 1 and Type 2 baskets may be stored in the HSM-HS storage module which is designed for locations with higher seismic levels. Based on a review of the information in the application, the staff finds that the applicant adequately described the HSM changes to the 24PTH storage system.

1.1.4 Transfer Cask Changes The applicant neither introduced new transfer cask (TC) designs nor modified existing TC designs in this amendment. The applicant updated a table in FSAR section P.4.1 to identify the TC designs authorized for use with a 24PTH DSC utilizing the new Type 3 basket. The applicant also added text clarifying that the previously authorized OS200/OS200FC TC is only utilized with Type 1 and Type 2 baskets. In addition, the applicant clarified external air circulation is needed for 24PTH-S or 24PTH-L DSCs during the transfer mode for all basket types. Based on a review of the information in the application, the staff finds that the applicant adequately described the TC changes to the 24PTH storage system.

1.2 System Drawings The applicant provided drawings for both the new Type 3D basket, the 24PTH-S-LC DSC lead shielding design changes, and the FFC used in conjunction with the Type 3D basket in FSAR appendix P, chapter 1. These included drawings of the structures, systems and components (SSCs) important to safety (ITS). After reviewing the drawings, the staff finds that they contain sufficient detail on dimensions, materials, and specifications to allow for a thorough evaluation

7 of the NUHOMS 24PTHs. The staff analyzed specific SSCs in sections 3 through 12 of this SER.

1.3 Technical Qualifications of Applicant FSAR appendix P, section 1.3 identifies agents and contractors. TN updated their company name in this section from Transnuclear, Inc. to TN Americas, LLC. The staff considered this change to be editorial; therefore, the staff finds it acceptable.

1.4 Changes to Standardized NUHOMS Certificate of Compliance A full list of CoC changes and detail regarding each change is provided in chapter 15 of this SER.

1.5 Changes to Standardized NUHOMS CoC Appendices A full list of changes to CoC appendices, and detail regarding each change, is provided in chapter 15 of this SER.

1.7 Evaluation Findings

The staff reviews the FSAR Chapter 1, General Description, to ensure that the applicant has provided a non-proprietary description, or overview, in its documentation for the spent fuel storage system that is adequate to familiarize reviewers and other interested parties with the pertinent features of the system.

The staff determined that the proposed description in general information is adequate for the NRC staff to conduct its evaluation as documented in the following sections of this SER.

Therefore, the staff concludes that the general description presented in the FSAR satisfies the requirements of 10 CFR Part 72. This finding is reached based on a review that considered the applicable regulations, applicable regulatory guides, and accepted practices.

F1.1 A general description of the Standardized NUHOMS 24PTH DSC system is presented in appendix P of the FSAR, with special attention to design and operating characteristics, unusual or novel design features and principal considerations important to safety.

F1.2 Drawings for SSCs important to safety are presented in appendix P of the FSAR.

2.0 DRY STORAGE FACILITY SITE CHARACTERISTICS EVALUATION Because this application dealt with a dry storage system, the staff did not evaluate the characteristics for a dry storage facility located at a specific location.

3.0 PRINCIPAL DESIGN CRITERIA EVALUATION The staff reviewed the proposed changes in the application (UFSAR Section P.2) to ensure the principal design criteria related to structures, systems, and components (SSCs) important to safety (ITS) comply with the relevant general criteria established in the requirements in 10 CFR Part 72. The design criteria for the Type 3 basket (Change No. 1) is discussed in UFSAR Section P.2.2.5.1.3.

8 Based on the review that considered the applicable regulations, regulatory guides, codes and standards, and accepted engineering practices, the staff determined the proposed principal design criteria are acceptable as documented in the following sections of this SER (i.e., Section 4.2 and 8.1).

4.0 STRUCTURAL EVALUATION The staff reviewed the information provided by the applicant and found one change that required structural evaluation:

Change No. 1: Addition of the Type 3 basket to a 24PTH DSC in the Standardized NUHOMS HSM system.

This section evaluates the structural adequacy of the 24PTH DSC Type 3 basket design.

Structural design features are reviewed together with evaluations of the structural analyses performed by the applicant to demonstrate the structural performance of the basket under normal and accident conditions.

4.1

System Description

The applicant designed the Standardized NUHOMS horizontal cask system for storage of spent fuel assemblies (SFAs) in a DSC and placement in a concrete HSM. Various canister and basket configuration designs permit storage of intact and damaged Pressurized Water Reactor (PWR) and Boiling Water Reactor SFAs. The welded DSC shell assembly consists of a cylindrical shell, canister bottom/top cover plates, and shield plugs or shield plug assembly. The DSC shell assembly contains a basket assembly of compartments formed from steel plate structures to which aluminum plates, for heat dissipation, and neutron poison plates are attached. Additional transition rails form the interface for the radial basket exterior with the DSC shell.

The applicant designed the 24PTH DSC Type 3 basket for storage of up to 24 PWR spent fuel assemblies (with a maximum of 12 damaged PWR SFAs) and a total heat load up to 40.8 kW.

The 24PTH DSC has a cavity length of up to 175.1, an outside diameter of 67.19, and a dry loaded weight of 93.7 kips. The 24PTH DSC shell assembly has three configurations: 24PTH-S, 24PTH-L, and 24PTH-S-LC. The designs differ in the DSC shell nominal thicknesses, outer top cover plate, inner top cover plate, inner bottom cover plate, top shield plug, and bottom shield plug. The 24PTH DSC basket assembly has three basket options: Types 1, 2 and 3. The basket assemblies differ in design of the transition rails and basket plates as well as different basket poison plates with varying boron content.

The Type 3 basket utilizes an egg crate type structure with interlocking, slotted steel plates.

Therefore, the entire Type 3 basket assembly requires no welds. Bolts secure the aluminum R45 and R90 transition rails to the radial exterior of the basket, and tie rods connect the R90 transition rails to each other. The Type 3 basket assembly design allows it to be unattached to the 24PTH DSC shell in any way. The applicant presented its structural evaluation with model analysis for the Type 3 basket in SAR section P.3.8 (Reference 2).

4.2 Evaluations of the 24PTH DSC Type 3 Basket The applicant provided the structural evaluation of the 24PTH DSC in SAR section P.3 and the evaluation of the Type 3 basket in SAR section P.3.8. The applicant indicated that the weight

9 and heat load of the 24PTH DSC is bounded by the 32PTH1 DSC. The staff previously reviewed and accepted the 32PTH1 DSC (ADAMS Package Accession No. ML090400180). As a result, the required additional analyses focused on the unique egg-crate design of the Type 3 basket. The applicant evaluated the 24PTH DSC Type 3 basket under normal and accident conditions using finite element (FE) methods and engineering equations based on the theory of elasticity. For normal conditions, the applicant used ANSYS FE models to evaluate the basket with the combined thermal and handling loads for the DSC in its horizontal storage orientation.

The model simulated a cross-section of the DSC and basket representing the hottest and most flexible middle portion of the DSC. For accident conditions, the applicant used the same FE model to evaluate the side drop orientation. The applicant evaluated the side drop accident with and without inclusion of bolts and the transition rail tie rods. For the accident condition end drop orientation, the applicant used a hand calculation to perform the analysis. The applicant also performed additional analyses to evaluate creep deformation and crack growth in the basket plates. In addition, the applicant evaluated the basket plate stresses, strains, and deformations against allowable values to demonstrate that the Type 3 basket in the 24PTH adequately performs its criticality safety function.

4.2.1 Evaluation of ANSYS FE Model Analysis SAR section P.3.8 described the ANSYS FE model of the 24PTH DSC Type 3 basket for side loading and impact and SAR figure P.3.8-1 provided an isometric view of the model. The applicant developed the ANSYS model analysis to investigate if deformation or collapse of the basket compartments would adversely affect the DSC criticality analysis assumptions. The model included representations of the DSC shell, steel grid plates, R45 and R90 aluminum transition rails, steel angle rail reinforcement, bolts, and tie rods. Although the model did not explicitly include the SFAs, the model simulated their influence by applying pressure to the basket compartment walls based on the magnitude and direction of acceleration loading considered in SAR figure P.3.8-3. The model ignored the aluminum basket plates and poison plates for conservatism, but the model included the mass of these components in the mass of the steel grid plates. This approximation neither contributed any structural support to nor resisted deformation of the steel grid plates. The applicant solved the quasi-static model for a given acceleration loading for normal and accident conditions.

4.2.1.1 Geometry The ANSYS model geometry explicitly represented the DSC shell, egg-crate steel basket plates, aluminum transition rails, steel angle rail reinforcement, bolts, and tie rods for the cross-section at the middle of the DSC. Screws attach the R90 aluminum transition rails to the basket, and tie rods attach the rails to each other. Screws attach the R45 aluminum transition rails and steel angle supports to the basket.

The central grid structure consists of a steel grid plate, an aluminum plate, and a MMC poison plate. The other grid structures utilize the same components; however, the other grid structures use a thinner aluminum plate versus the central grid structure. The outside of the basket consists of steel grid plates only. The DSC shell is 0.5 thick, which is a reduction from the previous 0.625 thickness of the 24P DSC, with a 66.19 inner diameter. The basket compartments have a nominal 8.9 pitch. The basket design also includes aluminum R45 transition rails and R90 transition rails as well as steel R45 angle plates. Screws attach the R45 transition rails, the R90 transition rails and the R45 angle plates to the steel grid plates. Tie rods and washers hold the R90 transition rails together.

10 The staff reviewed the dimensions of these components in the ANSYS model and found them to be consistent with the drawings in SAR section P.1.5 and presentation in SAR section P.3.8.

4.2.1.2 Material Properties The applicant used elastic material properties for all the materials under normal conditions. The grid plates are coated high strength low alloy (HSLA) steel (e.g., AISI 4130), the transition rails are aluminum 6061, the screws and tie rods are ASTM A564 Type 630 H1100, the angle supports are ASTM A516 Grade 70, and the DSC shell is Type 304 stainless steel. Elastic modulus, Poissons ratio, and density were assigned consistent with the SAR. For the accident condition analysis, the applicant used elastic-plastic material properties for all the materials. The staff reviewed and confirmed that stress-strain curves used in the ANSYS structural model analysis are consistent with the stress-strain curves described in the SAR.

The applicant calculated the allowable stresses of the steel basket using the methodology of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC)

Section III Division 1 Subsection NG for Level A and Level D service loading conditions and provided the acceptance criteria in SAR table P.3.8-1. The applicant took the allowable stresses for the threaded fasteners from NG-3230 of ASME BPVC 2010 Edition thru 2011 Addenda and provided the acceptance criteria in SAR table P.3.8-2. SAR table P.3.8-3 identified the following temperatures for allowable stresses: 700°F for the steel grid plates and 550°F for the remaining components. These temperatures bounded the maximum temperature for off-normal conditions in a horizontal transfer cask.

SAR table P.3.8-4 defined the following allowable plastic strains for accident analysis of the HSLA basket plates: 1% for primary membrane, 3% for primary membrane plus bending, and 10% for primary plus peak strains. These allowable plastic strains ensure that displacement and deformation of the basket grid is small and within failure limits for HSLA materials. The staff reviewed the material properties used in the ANSYS Type 3 basket model and found them to be consistent with the material properties presented in the SAR.

4.2.1.3 Elements The ANSYS model used a variety of different elements to represent the basket plates, DSC shell, transition rails, transition tie rods, attachment screws, the force-deflection curve of the attachment screws and tie rod washers. Contact elements represented contact between the basket plates, transition rails, angle plates, DSC shell, screws, and tie rods. The staff reviewed the selection of the element types and mesh size to model the basket. The staff finds that the element selection and mesh density adequately represent the Type 3 basket in the 24PTH DSC for the FE analysis.

4.2.1.4 Boundary Conditions The applicant used displacement boundary conditions to simulate a central section of the DSC and basket. The model included the plate intersections composing the egg-crate structure using axial constraints applied at the centers of the horizontal and vertical grid plates. The applicant applied axial constraints to all nodes of the DSC shell, transition plates, and angle support plates on the front and back model surfaces. After reviewing the boundary conditions applied to the ANSYS FE basket model, the staff finds that the boundary conditions applied adequately represent the Type 3 basket performance.

11 4.2.1.5 Loads The applicant modeled normal condition side drops with 1g dead weight plus 1g handling loading using three handling orientations: 1g vertical, 1g transverse, or 0.5g vertical plus 0.5g transverse. A bounding pressure load of 11.5 lb./in., applied to the steel basket plates as shown in SAR figure P.3.8-3, modeled the inertial loads due to the SFAs. For the normal condition temperature, the model applied a proprietary temperature profile to the basket to determine the range of temperatures for basket components. The model also calculated thermal expansion stresses for the basket components. Because axial clearances are provided between the DSC and basket components, no differential thermal expansion stresses occurred with the Type 3 basket. A quasi-static ANSYS analysis evaluated the accident side drop of the fuel rod using a bounding deceleration of 75g. The ANSYS analysis also evaluated side drop conditions for five basket orientations relative to the drop acceleration: 0, 180, 210, 225, and 270 degrees.

4.2.1.6 ANSYS Model Analysis Results The applicant performed the structural analysis using the ANSYS FE model with the material properties, boundary conditions and loadings described above, and provided the following analytical results for normal and accident conditions:

The maximum stress ratio for structures and fasteners was 0.82 under normal conditions, where the stress ratio is defined as a numerical ratio of the calculated stress with respect to the allowable stress indicating a safety margin when the stress ratio is less than 1.0.

The 210° case proved to be the bounding side drop orientation under accident conditions as shown in SAR table P.3.8-6.

The plastic strain of the grid plates remained less than the defined allowable as shown in SAR table P.3.8-7.

The factor of safety for basket buckling exceeded 1.25, which is higher than the required 1.0 as shown in SAR tables P.3.8-8 and P.3.8-9, since the grid plates did not buckle from a 94g acceleration loading.

SAR table P.3.8-10 provided the following maximum relative displacements for adjacent fuel basket compartments: X=0.09 and Z=0.11.

SAR table P.3.8-11 provided the following maximum relative displacements based on the updated model for the bounding orientation of 210°: X=0.077 and Z=0.131.

The staff reviewed and evaluated the applicants design and analysis of the 24PTH DSC Type 3 basket. This included ANSYS structural models of the 24PTH DSC Type 3 basket developed by the applicant to support the safety basis for normal and accident conditions. For normal conditions, the analyses evaluated combined thermal and handling loads in a horizontal storage orientation. For accident conditions, the analyses evaluated the side drop orientation with and without inclusion of the bolts and tie rods.

Based on the review, the staff finds that the technical approach and ANSYS models developed for the structural analysis of the 24PTH DSC Type 3 basket are adequate and acceptable. The staff also finds that the analytical results provide a reasonable assurance of safety against basket deformation that could affect the package criticality analysis. The ANSYS modeling of

12 the DSC basket well represented the deflection and buckling of the 24 PTH basket under side loading, and the results of the DSC basket showed that (i) all calculated stresses are smaller than the allowable values, and (ii) deformations are small and will not affect the analysis of subcriticality.

4.2.2 Evaluation of End Drop Analysis Under Accident Condition The applicant calculated axial stresses in the plates using a linear elasticity equation based on Hooke's law to show the HSLA basket plates would not experience significant strains under end drop as discussed in SAR section P.3.8.7.4. The applicant calculated a maximum axial stress in the plates of 17.75 ksi, which is less than the minimum 66 ksi yield stress shown in SAR table P.3.3.-10, considering the basket weight only. Additionally, the applicant performed buckling calculations and eigenvalue analysis of the basket plates in accordance with NUREG-6322 to demonstrate that the axial stress in the plates was less than the calculated buckling stresses.

As a result, the applicant concluded that an end drop was not of concern for basket deformation under accident conditions. The staff reviewed the calculation and confirmed that basket deformation under the end drop accident condition on the HSLA plates of the Type 3 basket is not a concern because the calculated maximum axial stress of 17.75 ksi in the plates is significantly less than the minimum yield stress of 66 ksi and the minimum buckling stress of 32 ksi.

4.2.3 Evaluation of Creep Model Analysis for Long Term Storage The applicant stated that the aluminum R90 rails are designed to resist the bearing loads due to the deadweight of the loaded basket for 80 years while stored in the HSM. For that reason, the applicant considered long-term creep effects of loading on the aluminum transition rail, and performed an evaluation to show the aluminum transition rails would not experience excessive creep deformation as presented in SAR section P.3.8.6.6. The applicant calculated a bearing stress of 46 psi on the R90 rail during horizontal storage with 1g loading. As a result, the applicant concluded that creep deformation is not of concern. The staff reviewed the calculation and confirmed that basket deformation under long term storage conditions is not a concern because the calculated maximum aluminum bearing stress of 46 psi is significantly lower than the allowable creep stress of 758 psi.

4.2.4 Evaluation of Crack Propagation and Growth Analysis Using the formula provided in NUREG/CR-1815 (Reference 4), the applicant calculated a critical nominal stress required for brittle crack growth initiation. The applicant performed this calculation to show that the HSLA basket plates yield before brittle crack growth begins. The staff compared the calculated critical nominal stress for brittle crack growth initiation with the HSLA steel material ultimate stress of 105 ksi at -20°F shown in SAR table P.3.3-10. The staff determined that this calculated critical stress is significantly higher than the ultimate stress.

Based on the critical stress calculation and the minimum specified fracture toughness for general HSLA steel from CoC appendix A section 2.4, i.e., 150 ksi in at -40°F, for a 1/16 through-crack in a basket plate, the staff finds that the plate will deform prior to any crack growth and that brittle fracture will not be of concern for the basket.

4.3 Evaluation Findings

The staff concludes that the structural properties of the SSCs of the Standardized NUHOMS Horizontal Modular Storage System are in compliance with 10 CFR Part 72 and that the applicable design and acceptance criteria have been satisfied. The evaluation of the structural

13 properties provides reasonable assurance that the Standardized NUHOMS Horizontal Modular Storage System will allow safe storage of spent nuclear fuel (SNF) for a certified life of 40 years.

This finding is reached on the basis of a review that considered the applicable regulations, appropriate regulatory guides, applicable codes and standards, and accepted engineering practices. The key findings from the staffs review of Amendment No. 18 include:

F4.1 On the basis of the review of the statements and representations in the application, the staff finds that the application adequately describes the 24PTH DSC Type 3 basket to enable evaluations of its structural performance and effectiveness.

F4.2 The staff finds that the application has met the requirements of 10 CFR 72.236(b). The 24PTH DSC Type 3 basket is designed to accommodate the combined loads of normal and accident conditions with an adequate margin of safety. Stresses at various locations of the basket for various design loads are determined by analysis. Total stresses for the combined loads of normal and accident conditions are acceptable and are found to be within the limits given in applicable codes, standards, and specifications.

4.4 References

1. U.S. Nuclear Regulatory Commission (NRC), Certificate of Compliance No. 1004 for the Standardized NUHOMS Horizontal Modular Storage System, Amendment 17.

2. Orano TN, Application for Amendment 18 to Standardized NUHOMS Certificate of Compliance No. 1004 for Spent Fuel Storage Casks, Revision 0 (Docket No. 72-1004),

Letter E-60447, May 20, 2022.

3. Orano TN, Updated Final Safety Analysis Report for the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, Revision 20, August 2021.

4. U.S. Nuclear Regulatory Commission (NRC), NUREG/CR-1815, Recommendations for Protecting Against Failure by Brittle Fracture in Ferritic Steel Shipping Containers Up to Four Inches Thick, June 1981.

5.0 THERMAL EVALUATION The objective of the staffs review of the applicants thermal evaluation for Amendment No. 18 to the Standardized NUHOMS System was to verify that the cask and fuel material temperatures will remain within the range of allowable values or criteria for normal, off-normal, and accident conditions. Specifically, the staff analyzed whether the temperatures of the fuel cladding will meet regulatory requirements throughout the storage period and will protect the cladding against degradation that could lead to gross rupture.

The staff reviewed the information provided in the amendment request to determine whether the Standardized NUHOMS System continues to fulfill the acceptance criteria in chapter 5 of NUREG-2215. The following changes in the application that involve thermal consideration are described below:

Change No. 1: Provide for a lighter 24PTH Type 3 basket design for the 24PTH DSC.

Change No. 2: Delete the ITE appendix A requirement for the initial HSM delta T measurement with a loaded DSC.

14 Change No. 7: Clarify in TS appendix B LCO 3.1.3 that there is no transfer time limit associated with the 24PTH-S-LC DSC, consistent with the existing UFSAR analysis.

This SER section documents the staffs review and conclusions with respect to thermal safety.

5.1 Change No. 1 - Lighter 24PTH Type 3 Basket Design In section P.4.12 of the UFSAR, the applicant described the thermal analysis for the 24PTH DSC Type 3 basket design for heat load zoning configuration (HLZC) #1 through #6 for storage and transfer conditions. The applicants thermal evaluations sought to demonstrate that the thermal performance of the Type 3 (also referred to as Type 3D in the UFSAR and TS) basket, which was proposed for use in the 24PTH-S, 24PTH-L, and the 24PTH-S-LC DSCs, exceeds the performance of the Type 1 and 2 baskets. The applicant performed sensitivity analyses that were based on storing the maximum heat load for the 24 PTH system, 40.8 kW, in either 24PTH-S or 24PTH-L DSCs as shown in the table on UFSAR page P.1-2. The applicant stated on UFSAR page P.4-2 that neither transfer operations in the OS200FC / OS200 nor storage in the HSM-HS are permitted with the Type 3 basket. As summarized on UFSAR page 4.1-2, the applicant only permits use of the Type 3D basket with the OS197 TC, and as shown in the table on UFSAR page P.1-2, with the HSM (model 102 or 202) and the HSM-H.

5.1.1 Storage Analysis of the 24PTH DSC with the Type 3 Basket In UFSAR section P.4.12.1, the applicant described the storage analysis of the 24PTH DSC with the Type 3 basket in the HSM-H. The description included the bounding storage condition, material properties, effective thermal properties of composite basket plates, bounding effective thermal properties of PWR FAs loaded in the 24PTH DSC. The discussion also encompassed the computational fluid dynamics (CFD) methodology, the CFD model description, the CFD meshing, heat generation in the fuel assemblies, boundary conditions, grid convergence index (GCI) calculation, maximum component temperatures, and maximum internal pressures.

The half-symmetric thermal model used by the applicant included conduction within the basket, radiative heat transfer between the DSC shell, heat shields, and HSM-H, and convection within the HSM-H cavity. In addition, the thermal model depicted conduction within the HSM-H structure as well as convection and radiation from both the HSM-H surfaces and vent outlet to the ambient environment. UFSAR table P.4-54 included results that showed the maximum component temperatures for the 24PTH DSC, with the Type 3 basket, and the HSM-H with HLZC#1 under normal storage conditions are bounded by the design basis temperature values in UFSAR tables P.4-14 and P.4-15. Because of this, the applicant also described that the off-normal storage and accident blocked vent conditions are also bounded for the 24PTH DSC with the Type 3 basket. The applicant calculated the GCI and also performed a sensitivity study by increasing the gap size between the basket and the DSC shell; the fuel cladding temperature results remained bounded by the design basis temperature value. The applicant did not include solar insolation in the analysis; however, a sensitivity study submitted by the applicant for EOS Amendment No. 1 demonstrated that insolation had an insignificant effect on the thermal performance of the EOS HSM-MX. The staff found this approach acceptable for EOS Amendment No. 1 as described in section 4.2.4.2.1.1 of the SER issued with Amendment No. 1 (ADAMS Accession No. ML20136A052).

The staff concludes that there would also be an insignificant effect on the thermal performance of the Standardized NUHOMS-24PTH DSC with the Type 3 basket within the HSM or HSM-H systems due to similarities between the EOS storage system and the Standardized NUHOMS

15 storage system, e.g., the thermal inertia of the systems loaded with a DSC. The applicant also explained that the maximum internal pressures in UFSAR tables P.4-19, P.4-24, and P.4-29 remain bounding for the 24PTH DSC with the Type 3 basket under normal, off-normal, and accident storage conditions, respectively. The staff finds this explanation acceptable because the average DSC cavity gas temperatures remain bounded by the design basis values that are discussed within UFSAR section P.4.6.5.4.

5.1.2 Transfer Analysis of the 24PTH DSC with the Type 3 Basket The applicant described the transfer analysis of the 24PTH DSC with the Type 3 basket in the OS197 TC in UFSAR section P.4.12.2. The description included the bounding transfer condition, material properties, CFD model description, CFD meshing, heat generation in the fuel assemblies, boundary conditions, GCI calculation, maximum component temperatures, and applicable time limits. UFSAR table P.4-64 included results that showed the maximum fuel cladding and component temperatures for the 24PTH DSC, with the Type 3 basket in the OS197FC TC, are lower than the associated temperature limits in UFSAR table P.4-64 during various conditions of transfer operations. The applicant calculated the GCI and continued to demonstrate with the consideration of that calculation that the fuel cladding temperature results remained bounded by the design basis temperature value.

The staff reviewed UFSAR section P.4.12.2.5.3 that discussed the rationale for the 9.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months time limit for the OS197FC TC loaded with the 24PTH DSC with the Type 3 basket for HLZCs

1, #2, #3, or #6. The applicant also described in UFSAR section P.4.12.2.5.3 that there is no time limit for the OS197FC TC loaded with the 24PTH DSC with the Type 3 basket with HLZC

4 with a decay heat of 31.2 kW or less. The applicant also described for the 24PTHS and the 24PTHL DSCs with the Type 3 basket, if the required time limit for completion of a DSC transfer is not met and the TC is in a horizontal orientation on the transfer skid, air circulation in the TC/DSC annulus is initiated by running the blower for a minimum of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Then, after the blower is turned off, the time limit for completion of DSC transfer is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , which is based on the maximum allowable heat load of 40.8 kW. The staff confirmed that the time limits described in UFSAR section P.4.12.2.5.3 were based on the results provided in UFSAR figure P.4-72 and were also included in LCO 3.1.3 of the Standardized NUHOMS CoC No. 1004. The staff finds the transfer analysis of the 24PTH DSC with the Type 3 basket and the associated transfer time limits to be acceptable based on the staffs review described in section 5.1.2 of this SER.

5.1.3 Thermal Model Audit As part of the staffs thermal model audit of the applicants CFD models, the staff reviewed the applicants thermal models used in the analyses and confirmed 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 prevalent in the Standardized NUHOMS CoC No. 1004 with the 24PTH DSC with the Type 3 basket that is within the HSM and OS197 TC for the geometry and analyzed conditions.

5.2 Change No. 2 - Delete Appendix A Requirement for Initial HSM Delta T Measurement with Loaded DSC The applicant proposes to delete section 4.4, HSM Maximum Air Exit Temperature with a Loaded DSC, from the Standardized NUHOMS CoC No. 1004 Amendment No. 17, appendix A (ML21109A329). This section required the applicant to verify that the HSM maximum air temperature rise limit is satisfied after the initial placement of a loaded DSC within the HSM.

16 In support of this change, the applicant described in the response to supplemental information request 4.1 (ADAMS Accession No. ML22224A041) that:

The same temperature measurements have been taken on hundreds of Standardized NUHOMS CoC No. 1004 systems after loading.

The procurement and fabrication processes are controlled by an approved quality assurance program.

The small tolerances permitted during fabrication for these components will not create a measurable impact on the air temperature measurement.

The heat load of the fuel assemblies is developed based on approved quality assurance programs and appropriate candidates are selected for loading into a DSC by the general licensee. There are numerous administrative controls throughout the entire process that ensure the correct fuel assemblies are identified and loaded into the DSC.

The applicant asserted that the controls used during the fabrication of the systems and the heat load determination for a given DSC made the air temperature rise measurement in the Standardized NUHOMS CoC No. 1004 appendix A section 4.4 redundant since the same measurements have been taken for hundreds of systems. The applicant noted that the HSM thermal monitoring program governed by the Standardized NUHOMS CoC No. 1004 appendix B TS 4.3.6, where the maximum allowed air temperature rise through the HSM as a function of the DSC decay heat load and the HSM model is located, would continue to ensure the thermal performance. The applicant concluded in its response to request for supplemental information 4.1 (ADAMS Accession No. ML22224A041) that the surveillance activities (visual or temperature monitoring) required per the HSM thermal monitoring program will ensure that licensees identify any potential conditions that could result in a blocked vent condition and that licensees will take subsequent corrective actions to clear the blockage.

Based on the staffs review of the applicants description of the controls used during the fabrications of the systems and the determination of the heat load for a given DSC, as well as the hundreds of prior Standardized NUHOMS CoC No. 1004 system temperature measurements taken after loading, the staff finds the removal of appendix A section 4.4 to be acceptable because it is unnecessary to ensure thermal performance given the statements in support of the removal.

5.3 Change No. 7 - Clarify 24PTH-S-LC DSC Transfer Time Limit in TS Appendix B LCO 3.1.3, Consistent with the Existing UFSAR Analysis The staff reviewed the proposed thermal TS changes in appendix B LCO 3.1.3 and confirmed that there is no transfer time limit associated with the 24PTH-S-LC DSC. Based on this review, the staff finds that the proposed thermal TS changes are acceptable because they are consistent with the existing UFSAR analysis.

5.4 Evaluation Findings

F5.1 SSCs important to safety are described in sufficient detail in the UFSAR to enable an evaluation of their thermal effectiveness in accordance with 10 CFR 72.236(f) and 10 CFR 72.236(h). Storage container SSCs important to safety remain within their operating temperature ranges in accordance with 10 CFR 72.236(a) and 10 CFR 72.236(b).

F5.2 The Standardized NUHOMS-24PTH DSC with the Type 3 basket, as amended, within the HSM or HSM-H systems is designed with a heat-removal capability, verifiably and

17 reliably consistent with its importance to safety. The storage container is designed to provide adequate heat-removal capacity without active cooling systems in accordance with 10 CFR 72.236(f).

F5.3 The SNF cladding is protected against degradation leading to gross ruptures under normal conditions by maintaining the cladding temperature for 40 years below 400 °C (752 °F) in a helium gas environment. Protection of the cladding against degradation is expected to allow ready retrieval of the SNF for further processing or disposal in accordance with 10 CFR 72.236(g), 10 CFR 72.236(l), and 10 CFR 72.236(m).

F5.4 The SNF cladding is protected against degradation leading to gross ruptures under off-normal and accident conditions by maintaining the cladding temperature below 570 °C (1,058 °F) in a helium gas environment. Protection of the cladding against degradation is expected to allow ready retrieval of spent fuel for further processing or disposal in accordance with 10 CFR 72.236(g), 10 CFR 72.236(l), and 10 CFR 72.236(m).

F5.5 The staff concludes that the thermal design of the Standardized NUHOMS-24PTH DSC Type 3 basket, as amended, within the HSM or HSM-H systems is in compliance with 10 CFR Part 72, and that the applicable design and acceptance criteria have been satisfied. The evaluation of the thermal design provides reasonable assurance that the Standardized NUHOMS-24PTH DSC Type 3 basket, as amended, within the HSM or HSM-H systems will allow safe storage of SNF for a licensed (certified) life of 40 years.

This conclusion is reached on the basis of a review that considered the applicable regulations, appropriate regulatory guides, applicable codes and standards, and accepted engineering practices.

6.0 SHIELDING EVALUATION The purpose of the shielding review is to ensure that the design features relied on for shielding provide adequate protection against direct radiation from the storage systems contents. This includes ensuring adequate protection for operating staff and members of the public so that the total doses remain within applicable regulatory requirements during design-basis normal operating, off-normal, and accident conditions. The shielding review seeks to ensure the shielding design is adequately defined and evaluated to support evaluation of the storage systems compliance with 10 CFR 72.236(d) and evaluation of the occupational doses from operations, including adequate consideration of as low as is reasonably achievable (ALARA) in the system design and operations.

For this amendment, the applicant proposed the following changes that are relevant to shielding:

Change No.1: a modified basket design for the 24PTH DSC variants. As part of this change, the applicant increased the maximum allowed weight for the contents (fuel assembly and control components). As part of this change, the applicant modified the lead in both shield plugs of the 24PTH-S-LC DSC to allow for the use of precast lead as an option to the poured lead in the currently approved design. The modification includes differences in lead density and thickness and the introduction of radial gaps between the lead and the steel components of the shield plugs. For this change there were also associated changes to the dose rate limits in sections 3.2 and 3.3.2 of appendix A to the CoC. The staff also identified that other differences were proposed for application of dose rate limits to the different 24PTH DSC variants in the TC and the HSM.

18 Change No. 9: incorporation of administrative controls during short duration operations that are not analyzed for tornado hazards. The aim of instituting the administrative controls is to ensure against an accident resulting from a tornado hazard that challenges the bases for determining the system meets the radiation protection requirements in 10 CFR Part 72, including not exceeding the regulatory dose limits in 10 CFR Part 72, by putting the storage system in a configuration that is outside the bounds of the shielding analysis. The adequacy of these administrative controls for short duration operations is evaluated in SER section 11. The adequacy of these controls is a basis for the staffs finding of adequacy of the shielding analyses.

Additional Scope Change No. 1: The applicant also requested, as an additional scope item, a change to the HSM concrete, to allow use of different cement, which is a blended Portland cement meeting the requirements of the ASTM C595 standard.

6.1 Modified 24PTH Basket and 24PTH-S-LC DSC Shield Plug Lead 6.1.1 Shielding Design Description The applicant proposed a new basket type for the 24PTH DSCs. This basket type, referred to as the Type 3 basket, differs significantly in construction from the two existing basket types for the 24PTH DSCs and will be used with all three 24PTH DSC variants, i.e., the 24PTH-L, 24PTH-S, and 24PTH-S-LC. One difference is that the Type 3 basket uses high strength low alloy (HSLA) steels that meet the new specifications that the applicant proposed to be added to CoC appendix A section 2.4. The manner of basket construction also differs from the two existing basket types. Although the basket cell walls continue to be a layered configuration, the layers no longer form a sandwich of aluminum alloy plate and neutron absorber plate between steel plates. The new configuration has steel plate on one side of the neutron absorber plate and aluminum alloy plate on the other side. The steel and absorber plate thickness are constant throughout the basket cross-section while the aluminum alloy plate thickness varies depending on basket cell location. Some outer basket walls use slightly thicker steel plate.

The staff reviewed the basket drawings to ensure the drawings include the information necessary to both understand and describe the new Type 3 basket as well as to confirm that the basket design is adequately and appropriately evaluated in the shielding analysis. The staff identified that the drawings include nominal dimensions for the thicknesses of all the basket components credited in the shielding analysis except for the R90 transition rails. The staff notes that nominal dimensions, as opposed to simply reference dimensions, are important to ensure the dimensions are treated appropriately given their importance to the analyses and to the definition, or description, of the basket. Although the basket drawing did not include nominal dimensions for the R90 transition rails, the staff finds that there is still enough information in the drawings to determine a nominal thickness for the R90 transition rails.

Additionally, the drawings only identify reference dimensions for the basket and basket component heights, i.e., lengths. The staff expected that these dimensions should also have been provided as nominal dimensions and not reference dimensions because the axial height of the basket relative to the spent fuel and control component contents in the basket can have significant impacts on dose rates, particularly in locations close to the TC if the basket does not cover the entire axial length of the spent fuel and control component contents.

However, in reviewing the drawings, the staff determined that the Type 3 basket axial dimensions are the same as for the other basket types for the respective 24PTH DSC variant. In addition, the staff identified that the drawings contain a note about clearances between the basket and DSC components for meeting the thermal and structural performance requirements.

19 The staff determined that this note will also ensure the clearances between the basket and the top DSC shield plug are quite small which will have the effect of ensuring the Type 3 basket covers the full axial length of the spent fuel and control component contents.

The staff also investigated the dimensional and material tolerances of the basket components which are another important design aspect. For the shielding analysis, the applicant treated the aluminum alloy and neutron absorber plates as aluminum. This ignored any potential impact, though small, of the alloying elements and boron in these components and made material tolerances on these items unnecessary. In addition, because the steel must meet the industry material standard specifications identified in both the drawings and in section 2.4 of appendix A to the CoC, any material tolerances, e.g., density variation, will be captured by those specifications. The applicant did not specify explicit dimensional tolerances for the basket components. However, with all but the neutron absorber plates having an industry material standard specified for them, and with those standards including dimensional tolerances either directly or by reference to other standards that provide specifications on allowable dimensional variations, the staff finds the drawings include the necessary information to determine the basket components tolerances. The applicant indicated that the basket components tolerances are customized tolerances (ADAMS Accession No. ML23074A342); however, the staff noted that the tolerances described by the applicant align with tolerances the staff identified from the industry standards applicable to the basket components. The staff expects the tolerances in the standards will be met and that drawing tolerance changes will undergo the necessary 10 CFR 72.48 evaluations. The staff performed some simple calculations to estimate the impact of basket component tolerances on dose rates. Since the drawings do not specify tolerances for the neutron absorber plates, the staff used what it determined would be reasonable tolerances for them. The staffs evaluation indicated that the impact on dose rates from basket dimensional tolerances will be small and likely within the relative error of the calculation. Based on this evaluation, the staff finds the basket tolerances provided and determined by the means described above, and a lack of neutron absorber plate tolerances, to be acceptable for shielding.

The applicant also proposed to modify the lead shielding in the bottom and top DSC shield plugs in the 24PTH-S-LC DSC. This variant is the only variant of the three 24PTH DSC variants that has lead shield plugs; the other two variants have all steel shield plugs. The current 24PTH-S-LC DSC shield plug design uses poured lead. The applicant proposed a new option to use precast lead inserts that are machined to fit into the shield plug lead cavities. A correct understanding and definition of the allowed radial and axial gaps between the lead and steel shield plug components is important to both the shielding analysis and the shielding performance of the DSC. This includes for accident conditions where a drop accident may result in lead slump. The allowed gaps and gap sizes determine the maximum possible amount of lead slump in a drop accident and whether or not portions of the DSC contents may no longer be shielded by the lead in such an accident. Thus, the staff requested additional information about the allowed gaps and gap sizes for the lead inserts as this information determines the maximum possible lead slump in a drop accident and whether or not portions of the DSC contents may not be shielded by the lead in an accident. For this lead shielding option, the applicant specified maximum sizes of radial gaps between the lead and steel components of the shield plugs. The applicant specified that no axial gaps are allowed between the lead and the steel components of the shield plugs and revised the drawings to clearly preclude axial gaps.

While other analyses, e.g., the thermal analyses, include axial gaps to yield conservative results for those analyses, the applicant introduced no such axial gaps in the design.

The drawings provide a material density specification slightly less than the theoretical density of lead for the new lead shielding option. The drawings also identify the minimum axial lead

20 thicknesses in the shield plugs for the new lead shielding option. The minimum axial thickness of the new lead shielding option for the bottom shield plug is less than the minimum thickness of the poured lead option for the bottom shield plug. This lead thickness difference has an accompanying change in minimum axial steel thickness for the bottom shield plug that is also shown in the drawings. The applicant revised the drawings to ensure the lead thickness specifications are clearly understood to be minimum lead thicknesses for the fully fabricated DSC, as opposed to a minimum thickness at some intermediate stage such as prior to installation of the lead insert into its cavity. Thus, the staff finds the drawings adequately describe the new alternative DSC shield plug lead option.

Because the structural performance of the basket and the lead impacts the DSCs shielding performance, the staff sought to understand how the basket and lead changes impacted the analyzed structural performance of the DSCs. SER section 4 describes the structural review of the changes, and SER section 8 describes the materials review of the changes. The staff considered the results of those reviews in evaluating the shielding design and performance of the DSCs with the proposed modifications.

The applicant provided drawings for basket cell end caps for locations containing damaged fuel in the Type 3 basket and for failed fuel cans for this basket. After reviewing these drawings, the staff finds that they adequately describe aspects of the end caps and failed fuel can that impact the DSC shielding design and performance.

6.1.2 Radiation Source Definition The applicant did not propose any changes to the contents specifications that affect the radiation sources for the contents. The staff confirmed that there are no changes to the contents specifications in appendix B to the CoC for the 24PTH DSCs except to ensure a clear and correct understanding that the relevant burnup, enrichment, and cooling time specifications for the spent fuel contents are the maximum assembly average burnup, the minimum assembly average enrichment, and the minimum cooling time. The applicant modified the contents specifications for other NUHOMS system DSCs, as needed, to ensure the same specifications for those DSCs are also clear and correct.

The applicant has designed the Type 3 basket to be lighter in weight compared to the existing Type 1 and Type 2 baskets. Thus, the applicant proposed to increase the maximum weight of the spent fuel and control component contents allowed in each basket cell by 33 lbs.

(approximately 15 kilograms) and make conforming changes in table 1-1l of appendix B to the CoC. Although the applicant did not change the fuel source term specifications or increase the control components total source term, the change concerned the staff because the source term axial distribution could change depending on where the increased weight, and the accompanying radiation source due to hardware irradiation, was located which in turn could affect dose rates. While the extra mass in a particular axial location would provide some additional self-shielding, the staffs evaluation indicated that dose rates could increase significantly because the self-shielding from the increased mass may not compensate for any increased radiation source that might accompany the increased mass in that location. The applicant explained that the difference in allowed total contents weight accounts for the most conservative assembly and control component weights as well uncertainties in those weights, but the applicant also clarified that this did not change the axial distribution of the contents weight and the associated radiation source (ADAMS Accession Nos. ML23074A342 and ML23191A075). Thus, because both the total radiation source term, and the axial source term distribution, i.e., the source strength in each axial zone of the assembly and control components, remain the same as previously analyzed, the staff finds that there should be no

21 shielding impact and no increase in dose rates from the proposed change in the contents weight limit.

The contents specifications regarding damaged fuel also remain unchanged except for the point in table 1-1l of appendix B to the CoC regarding retrievability. This change is relevant to understanding the damaged fuel configuration under various conditions and the discussion in SER section 6.1.3.1 regarding the adequacy of the applicants shielding analysis to address damaged fuel contents. Since the change provides clarity to understanding that configuration and supports the sufficiency of the shielding analysis, the staff finds the change to be acceptable in terms of the DSC shielding performance.

6.1.3 Shielding Model Specification 6.1.3.1 Shielding and Source Configuration The applicant used the same shielding configurations to analyze the 24PTH DSCs with the Type 3 basket as it used with the previously approved basket types except that the basket specifications in the current model match the Type 3 basket. For the 24PTH-S-LC, the model also incorporated the additional difference of the alternate precast lead shielding insert in the DSC shield plugs versus the poured lead shielding. The applicant only performed new calculations for the different operating configurations under normal conditions, e.g., transfer, decontamination, and welding. Because the 24PTH-L bounds the 24PTH-S, the applicant only evaluated the 24PTH-L in the OS197FC TC and the 24PTH-S-LC in the Standardized TC. The staff finds this approach acceptable as these models explicitly capture the basket changes for the 24PTH DSCs and the DSC shield plugs lead shielding changes for the 24PTH-S-LC.

Instead of creating new models for the DSCs in the HSMs, the applicant used the ratio of the axial gamma dose rates for the DSCs in their respective TCs to adjust the relevant HSM dose rates which the applicant determined was the HSM door centerline. The applicant only scaled the HSM door centerline dose rates because its TC analyses indicated that the other dose rates would not change or would decrease. The staff finds this approach for HSM dose analysis acceptable because the staff performed simple confirmatory analyses indicating that relative differences in dose rates at the base of the TCs would be comparable to the relative differences in dose rates at the HSM doors. SER Section 6.1.4.3 contains further discussion regarding the dose rate trends at other locations around the TCs and HSMs.

The applicant did not create new accident condition models nor did it perform any new accident condition analyses because the applicant concluded that the current accident analyses remain bounding. For the HSMs, the applicant had previously shown that accident condition dose rates, except as noted below, are the same as for normal conditions for two reasons. First, the accident conditions analyses showed no impacts to the DSCs due to accident conditions.

Second, the accident conditions analysis showed that these conditions typically result in either no or minimal impacts to the HSMs shielding ability, particularly in terms of dose versus the 10 CFR 72.106 limits, except for HSM versions with a gap between adjacent HSMs that may widen in a seismic event. As discussed below, differences in TC radial dose rates for the Type 3 basket type versus the existing basket types indicate the existing accident analysis for the instance of wider gaps between HSMs will remain bounding. For the TCs, the applicants analyses assumed the radial neutron shielding is lost and that radial dose rates dominate the accident conditions. Since the axial shielding does not change from normal conditions to accident conditions, the axial dose rates remain the same under normal and accident conditions. The applicants analysis showed that normal condition radial dose rates for TCs containing DSCs with the Type 3 basket are lower than the radial TC dose rates for the current

22 basket types. Therefore, the applicant concluded that the current accident analysis remains bounding.

Based on its simple confirmatory analysis, the staff finds that TC radial dose rates for a DSC with the current basket types will bound the TC radial dose rates for a DSC with the Type 3 basket. The staff noticed that the TC radial accident dose rates are greater than the TC axial dose rates for a TC loaded with a 24PTH-L DSC, but that this is not the case for the 24PTH-S-LC in the TC. However, the staff finds that TC radial accident dose rates for the 24PTH-L will bound the TC axial accident dose rates for the 24PTH-S-LC.

In addition to comparing radial and axial TC dose rates, including comparing dose rates between TCs with the Type 3 basket and the current basket types, the staff considered that the shielding and dose rate evaluations should account for the behavior of damaged fuel under normal, off-normal, and accident conditions. For normal and off-normal conditions, the damaged fuel requirements in table 1-1l of appendix B to the CoC, as modified by the applicant in response to the staffs concerns and questions, provide assurance that damaged fuel will not reconfigure. In particular, the requirements limit the damage such that the assembly can be handled by normal means and ensure retrievability following normal and off-normal conditions.

Implementation of the requirement for retrievability is independent of the basket cell end caps.

For accident conditions, the applicant pointed to the reconfigured damaged fuel analysis for the 32PTH1 DSC. The staff noted that the 32PTH1 DSC analysis only considered the impacts on radial dose rates due to fuel reconfiguration. After comparing the 32PTH1 versus the 24PTH in terms of the damaged fuel loading configuration and other shielding-relevant parameters, the staff finds the use of the 32PTH1 analysis acceptable to justify that the 24PTH radial accident doses for intact fuel bound those for when the damaged fuel is reconfigured. The staff reviewed differences in axial dose rates versus the accident radial dose rates. Even if the axial dose rates doubled due to fuel reconfiguration, which the staff expects to be a bounding estimate for dose rate increases due to fuel reconfiguration, the TC accident radial dose rates with the 24PTH-L would still bound the TC accident axial dose rates with the 24PTH-L.

The staff questioned if lead slump due to a TC drop accident would impact the 24PTH-S-LC axial dose rates. The staff evaluated two factors of importance in the DSC lead insert shield plugs: any significant voids in the lead and the presence of allowable gaps as well as their sizes.

Based on the fabrication method for precast lead significantly limiting the existence and size of any voids in the lead, as explained by the applicant, and that only radial gaps are allowed, as specified in the design drawings, the staff evaluated the impacts of lead slump in both an axial drop and a horizontal drop. The staff estimated the potential increase in axial dose rates due to an accident condition axial drop, which would cause the lead to thin and expand to fill the lead cavitys cross-section. The staff performed this estimate with a simple, conservative hand calculation which evaluated the change in half value thicknesses due to the lead slump. The results indicated that the TC radial accident dose rates remain bounding. For an accident condition horizontal drop, the staff estimated the size of the cross-section which would have no lead between the DSC cavity and the DSC outer axial surface. For the maximum specified radial gaps, the staff determined that the size of this area will be small enough to prevent exposing any fuel basket cells, i.e., there is no direct streaming path for any of the DSC spent fuel contents. The staff also considered that the streaming path would have a negligible impact on axial dose rates at relatively short distances from the TC due to its small size.

Thus, the staff finds the applicants approach of not doing separate accident models and analysis to be acceptable given that the TC radial accident dose rates with the 24PTH-L DSC would bound all radial and axial dose rates even when considering reconfigured damaged fuel, lead slump due to an accident condition axial drop and the negligible impact on accident doses

23 from the small streaming paths caused by lead slump in accident condition horizontal drops.

Also, based on the damaged fuel evaluation noted above and the discussion regarding the increased allowable total assembly and control component weight not affecting the radiation source axial distribution, the staff finds the applicants use of the same source configurations as in the current analyses to be acceptable.

Additionally, because the DSC is stored horizontally in the HSM and 24PTH-S-LCs shield plugs design includes radial gaps between the shield plugs steel components and the precast lead inserts, the staff had questions about the potential for shielding and dose rate impacts due to lead creep while the DSC is stored in an HSM. Any potential lead creep could ultimately result in a lead configuration similar to that due to lead slump in a horizontal drop accident, with some part of the DSC cavity cross-section no longer shielded by the lead in the DSC shield plugs. Thus, the staffs evaluation regarding lead slump for the horizontal drop accident covers and would bound any concerns due to lead creep while the DSC is in the HSM. Furthermore, creep is a phenomenon that occurs over time during which the decay heat and radiation source terms of the contents are also decreasing at a rate that, as indicated by the applicant (ADAMS Accession No. ML23020A921), will likely outpace any lead creep. This consideration provides further support for the staffs conclusion that the impact of lead creep on shielding and dose rates would be small to negligible.

6.1.3.2 Material Properties The DSC contents material properties remained unchanged in the model. Based on considerations related to the contents, including those in the preceding sections, the staff finds this acceptable. The staff noticed that the applicant used the properties of stainless steel for the Type 3 basket HSLA steel. However, reference information available to the staff indicates that HSLA steels have a density that is lower than the density of both stainless steel and carbon steel. Thus, the staff considered the applicants model to be non-conservative in this regard.

The applicant provided the HSLA density in its design in UFSAR table P.3.3-10. This density resembles the densities used for carbon steel in previous shielding analyses that the staff has seen. The applicant indicated that the density in the UFSAR is only slightly less than the stainless steel density used in the model and would not have a notable impact on dose rates.

The staff evaluated the impact of the density difference on dose rates, including for the lower densities for the HSLA in the literature available to the staff, and found that the density difference has a very small impact on dose rates; therefore, the staff finds the applicants use of the stainless steel density acceptable.

The applicant used a lead density of 11.0 grams per cubic centimeter (g/cm3) for the new alternative DSC shield plug lead shielding. The drawings specify that the precast lead insert density must be 98.5 percent of leads full density, i.e., 11.34 g/cm3. This means the minimum density for the precast lead inserts must be 11.17 g/cm3. Therefore, the staff finds the use of 11.0 g/cm3 acceptable because it is less than the density specified on the drawings and so bounds the shielding performance of the lead.

6.1.4 Shielding Analyses 6.1.4.1 Computer Codes The applicant continued to use the same analysis techniques, including computer codes and code versions, that it used in previous analyses for the 24PTH DSCs as described in UFSAR section P.5.4.1. The staff determined there is nothing new or unique to the proposed changes

24 that make the use of those techniques and codes inappropriate. Therefore, the staff finds their continued use acceptable for this amendment.

6.1.4.2 Flux-to-Dose-Rate Conversion The applicant continued to use the same flux-to-dose-rate conversion factors for the analysis in this amendment that it has used in previous shielding analyses for amendments for this storage system. The staff finds these conversion factors to be acceptable because they are the same conversion factors for shielding analyses described in NUREG-2215 sections 6.4.4.2 and 6.5.4.2.

6.1.4.3 Dose Rates The applicant provided new dose rates based on its analyses. These new dose rates encompass the various operation configurations in the TC and HSM-H, an HSM version that is part of the NUHOMS storage system, for the 24PTH-L with a Type 3 basket. These configurations include the normal conditions decontamination, welding, and transfer operations in the TC and the normal storage conditions in the HSM-H. For the 24PTH-S-LC, the applicant provided new dose rates only for the normal conditions transfer operations in the TC and the normal storage conditions in the HSM Model 102, which is another HSM version that is also part of the NUHOMS storage system. The staff noted that the analyses for these DSCs also use different TC versions that are part of the NUHOMS storage system, which is consistent with the existing analyses for the 24PTH DSCs. As already noted, the applicant performed dose rate calculations for the DSCs in the TCs and used the ratio of gamma dose rates between the cases with the different baskets to provide a new door centerline dose rate for the HSMs with DSCs having the new Type 3 basket. The applicant did not generate any new TC accident dose rates because the normal conditions TC radial doses rates for the DSCs with the current basket types bound the TC radial dose rates for the DSCs with the new Type 3 basket and the applicants accident analysis focused on doses due to the radial side of the TCs.

The staff reviewed the dose rates and evaluated the differences in dose rates for the different basket types, and for the 24PTH-S-LC, for the changes in the bottom and top axial shield plugs lead shielding. As part of this review, the staff also conducted some simple calculations to determine what it would expect for differences in dose rates for these DSC design changes.

Based on these simple calculations, the staff confirmed that the radial dose rates should decrease when changing the basket to the new Type 3 basket.

The staff found that the axial dose rates should all increase, both at the top and bottom of the TCs, with greater increases occurring in the 24PTH-S-LC due to the lead shielding changes.

The staff anticipated that the most significant increases would occur at the TC base and at the HSM door centerline with the 24PTH-S-LC because the applicant made the most significant lead shielding changes to the 24PTH-S-LCs base shield plug and the DSC base faces the HSM door. In most instances, the trends in the applicants dose rate results proved to be consistent with the results of the staffs simple calculations although there were some notable differences.

These differences occurred for the TC top dose rates in the transfer configuration for both DSCs. The applicant reported only a small increase for the 24PTH-L and a slight decrease for the 24PTH-S-LC. The staff also expected that the relative increase in HSM door centerline dose rates for the 24PTH-S-LC would be substantially larger than the relative dose rate increase for the 24PTH-L; however, the applicants results showed essentially no difference. Additionally, the applicants response to the staffs questions regarding occupational dose estimates, e.g., table 10.1-1 of the response to additional information request 10.1 (ADAMS Accession No. ML23020A921) showed average dose rate changes more consistent with the staffs

25 calculations. This information seemed to further support the staffs concern regarding the dose rate results for the TC top.

Both in a clarification call and in supplemental information provided after the call, the applicant pointed to information in UFSAR chapter P.5, particularly figures P.5-20 and P.5-23, and provided additional information associated with the new calculations (ADAMS Accession Nos.

ML23074A342 and ML23191A075) that illustrate the dose rate profile across the TC top surface. This information indicated that the maximum TC top surface dose rates in the transfer configuration are located above the annulus between the DSC and the TC inner shell, which is impacted by radiation fluence changes on the radial side of the DSC due to the basket type changes. Since the Type 3 basket reduced the radiation levels on the DSC radial side, this reduced the axial dose rates above the annulus. The TC radial dose rate differences proved to be not insignificant. Thus, the top axial dose rate behavior, which is the maximum dose rate for the TC top in the transfer configuration, can be explained by the decrease in radial radiation levels with the new basket type competing with the increase in axial radiation levels through the top of the DSC. The staff evaluated the information provided by the applicant and finds it provides a reasonable explanation for the observed dose rate changes and why they differ from the staffs evaluation and expectations.

The staff noted that the applicants additional information showed the maximum dose rates for the TC top surface in the decontamination and welding configurations occur directly over the DSC lid. The staffs calculations applied to that same area and predicted a similar relative increase in dose rates there. The staff also noted that the applicants additional information showed changes in TC dose rates in the transfer configuration in areas directly above the DSC lid that are similar to changes identified in the staffs analysis for the 24PTH-L although they are somewhat lower than the staffs analysis predicts for the 24PTH-S-LC.

For the case regarding relative differences in the HSM door dose rates, the staff noted that the estimated relative increase for the 24PTH-L is significantly higher than what the staff expected whereas the increase for the 24PTH-S-LC is a bit lower than staff expected. In evaluating these dose rate trends; the staff considered several factors. First, the staff recognized and considered the limitations of the staffs simple calculation method. Further, the staff had no concerns with the applicants 24PTH-L results for the HSM door because they exceeded the staffs results.

While the staff predicted a greater increase for the HSM door dose rate with the 24PTH-S-LC versus what the applicant determined, the applicant specified a dose rate limit in the technical specifications at the door centerline that aligns with the applicants results for the 24PTH-S-LC in the HSM Model 102. These technical specification dose rate limits would preclude the dose rates estimated by the staff. Finally, the relative dose rate increase estimated by the staff would not have much of an impact on the HSM-H dose assessments because the door centerline dose rates for the 24PTH-S-LC in the HSM-H either are or would be quite small. Thus, the staff finds the dose rates from the new calculations to be reasonable and useful for the evaluations necessary to demonstrate compliance with 10 CFR 72.236(d) and the limits in 10 CFR 72.104 and 10 CFR 72.106.

The staff noted that the applicant did not include any revised results for the HSM average surface dose rates. While the top and side surfaces would have decreased dose rates due to facing the radial side of the DSC, the staff anticipated that the HSM front and rear surfaces would see increased dose rates due to the increased axial dose rates with the different basket and the different lead shield plug design. In particular, the staff focused on the front HSM surface. The applicant noted that, while the door centerline dose rates increased significantly, the front surface average dose rate is dominated by the birdscreen dose rates on the front vents and the door centerline dose rates make a relatively small contribution to the surface average.

26 The HSM configuration results in the vent dose rates being predominantly influenced by the DSC radial radiation levels which decreased with the new Type 3 basket. Thus, the applicant determined that the previously calculated average surface dose rates either would not change or would be bounding for the design changes to the 24PTH DSCs. The staff performed evaluations to estimate the dose contributions of the door and vents as well as to determine how changes to them would influence the HSM front surface average dose rates. The staffs evaluation indicated that there would be at most only nominal increases in average surface dose rates, and in some cases, average surface dose rates could potentially decrease. The staffs evaluation indicated that this would be true even when considering potential differences between evaluations by the staff and the applicant of the aspects described above that impact dose rates. Even in instances of possible dose rate increases, the increases proved to be small and minimally impacted the 10 CFR 72.104 dose evaluations. Thus, the staff finds the applicants analysis regarding the average surface dose rates acceptable.

6.1.4.4 Confirmatory Analyses The staff performed a variety of simple confirmatory calculations to evaluate the applicants amendment as noted in previous SER sections. The calculations included estimates of the impacts of dimensional tolerances, differences in material densities, differences in basket types, differences in DSC shield plug lead shielding, and evaluation of average surface dose rates for the HSMs. The staff used such methods as evaluating the half value thicknesses of materials credited for shielding at selected gamma energies and using differences in these thicknesses to estimate differences in dose rates. The staff also used MicroShield to provide estimates of relative dose rate differences. In conjunction with using the MicroShield code, the staff made simplifying assumptions and modeling choices due to code limitations. The staff remained cognizant of and factored in the limitations of these methods when using the results of these calculations to evaluate the applicants proposed changes and analyses.

6.1.5 CoC Appendix A 24PTH DSC Dose Rate Limits Dose rate limits in the CoC appendices serve a variety of functions. Effective limits have certain important characteristics. Section 6.6.1 of the SER for the initial approval of Amendment No. 11 of the NUHOMS storage system describes some of these functions and characteristics well (ADAMS Accession No. ML14010A486). The NUHOMS storage system includes dose rate limits for both the TC and the HSM in sections 3.2 and 3.3.2, respectively, of appendix A to the CoC that apply to the 24PTH DSCs. These sections also specify the dose rate measurement requirements for the TC and the HSM associated with each authorized NUHOMS system DSC, including the 24PTH DSCs, including the configurations for which the measurements are to be taken.

The staff reviewed the proposed changes to the dose rate limit values and found that they are supported by the applicants shielding analysis for the system configurations to which the limits apply. For some limits, the applicant included some margin versus the analysis results.

However, the staff determined that the margin is reasonable and not excessive for addressing measurement uncertainties. The clear connection between the analyzed dose rates and the dose rate limits, i.e., the limits are clearly based on and derived from the analyzed dose rates, is an important characteristics of effective dose rate limits.

The staff also noted that the applicant proposed changing the applicable limits for the 24PTH DSCs in another way. Amendment No. 17, section 3.2 of appendix A of the CoC explicitly identifies separate TC dose rate limits for the 24PTH-S-LC versus for the 24PTH-L or 24PTH-S (ADAMS Accession No. ML21109A329). Amendment No. 17, section 3.3.2 of appendix A of the

27 CoC also specifies the HSM dose rate limits such that only the 24PTH-L and 24PTH-S are allowed to be stored in the HSM-H while only the 24PTH-S-LC is allowed to be stored in the standardized HSM. In the proposed Amendment No. 18 CoC appendix A sections, the applicant combined the TC dose rate limits into a single set of limits applicable to all 24PTH DSCs. The applicant also revised the HSM-H limits to apply to the 24PTH-S-LC which would allow storage of that 24PTH variant in the HSM-H. The applicant did not propose a change to the restriction of only allowing storage of the 24PTH-S-LC in the standardized HSM.

In reviewing these proposed changes, the staff reviewed the history of the 24PTH DSC dose rate limits since the introduction of these DSCs into the NUHOMS storage system in Amendment No. 8. The staff also considered the shielding analyses for the 24PTH variants and the characteristics of these variants and their allowed contents and evaluated the proposed change against the discussion of the functions and characteristics of effective dose rate limits given in the Amendment No. 11 SER section stated above. The staff noted that, while the top axial TC limit differed for all three 24PTH variants prior to Amendment No. 11, the top axial TC limit has been the same for all three 24PTH variants since Amendment No. 11; however, the staff noted that there has always been a separate radial TC limit for the 24PTH-S-LC. In addition, the staff found that the HSM-H dose rate limits have often, but not always, limited the HSM-H to preclude storage of the 24PTH-S-LC and that the standardized HSM dose rate limits have always precluded storage of the 24PTH-S and 24PTH-L.

The staff had concerns with the efficacy of using the top axial TC limit, which is derived from the 24PTH-L analyses, for the 24PTH-S-LC. This concerned the staff because the TC axial top dose rates with the 24PTH-S-LC variant are much lower than for the 24PTH-L due to the differences in allowed contents specifications and the use of lead and steel shield plugs in the 24PTH-S-LC DSC versus steel only shield plugs in the 24PTH-L DSC with thicknesses that make the 24PTH-S-LC shield plug a more effective shield. However, the staff acknowledged the evaluation findings in regard to dose rate limits in the Amendment No. 11, Revision 0 SER, which applies the same axial top TC dose rate limit to all three 24PTH variants. Because nothing has changed since that time that would affect those findings as they apply to that particular limit, the staff finds applying the same axial top TC dose rate limit to all three 24PTH variants to still be acceptable. However, the staff finds that there should continue to be a separate radial TC limit for the 24PTH-S-LC to at least ensure that the radial TC dose rate limits facilitate meeting all the necessary functions and have effective limit characteristics, e.g., the limit is appropriate to the DSC and its contents without excessive margin. Thus, the applicant will retain the separate TC radial dose rate limit for the 24PTH-S-LC in section 3.2 of CoC appendix A.

In evaluating the HSM-H dose rate limits, the staff noted that the initial separate limits for the 24PTH-S-LC were intended to apply to this 24PTH variant in both the standardized HSM and the HSM-H. The UFSAR indicates that this 24PTH variant was intended be stored in both versions of the HSM. The shielding analyses for the 24PTH-S-LC used the HSM Model 102, a standardized HSM, because it has less shielding than the HSM-H. Thus, the dose rates for the 24PTH-S-LC in the HSM Model 102 bound the dose rates for the 24PTH-S-LC in the HSM-H.

The staff used simple calculations to estimate the dose rate impacts due to shielding differences between the HSM Model 102 and the HSM-H. Those estimates indicated that the end shield wall dose rate limit for the 24PTH in the HSM-H is a reasonable limit for the 24PTH-S-LC. The staffs simple calculations indicated that the HSM-H door limit would provide somewhat excess margin for the 24PTH-S-LC DSC; however, the staff finds the limit acceptable because the dose rate limit is low, i.e., approximately 5 mrem/hr, and other factors that impact dose rate measurements at these low radiation levels. The staff questioned if the HSM-H front birdscreen dose rate limit would be appropriate because the standardized HSM front birdscreen dose rate

28 limit is significantly below that limit and the calculated dose rates for the 24PTH-S-LC in the HSM Model 102 are below the limit as well. However, the staff finds the use of HSM-H front birdscreen dose rate limits adequate for storage of the 24PTH-S-LC because the applicant chose to retain the separate radial TC dose rate limits for the 24PTH-S-LC, which would be useful to identify any concerns with the DSC contents that would most affect the front vent dose rates, and because the staff expects both the applicant and licensees using the system will implement other measures to prevent problems with the fabricated HSM-H that could affect front vent dose rates. Thus, the staff finds it acceptable to apply the HSM-H limits for the 24PTH DSC to all 24PTH DSC variants.

Based on the foregoing evaluation, the staff finds the CoC appendix A dose rate limits for the 24PTH DSCs, as modified in this amendment request, acceptable.

6.2 HSM Concrete Cement Mixture Change The applicant proposed a change to the HSM concrete cement mixture. Instead of using Portland Type II cement that meets the requirements of ASTM C150, the applicant proposed transitioning to blended Portland cement meeting the requirements of ASTM C595. The staff evaluated the potential impacts of this change on the HSMs shielding performance, focusing on the potential impacts of the new mix properties on the concretes ability to shield gamma and neutron radiation as well as the ability to resist the effects of normal, off-normal, and accident conditions that could also degrade shielding capability. Based on the review findings in SER sections 4 and 8, the staff determined that the concretes structural and material performance as it relates to shielding will not differ from concrete using the cement mix that is currently part of the design.

In terms of differences in gamma and neutron radiation shielding capability, the staff performed some simple calculations to estimate the composition change impacts. To understand the impact, the staff reviewed the ASTM standard and other available information sources to understand what the concrete composition could be with the new cement mix as well as how much of the concrete composition is from the cement mix versus the aggregate. Based on this evaluation, the staff determined the dose rate impacts from the cement mixture change would be small. In addition, because the specified concrete density remains unchanged, this concrete property had no impact on dose rates. In sum, the staff finds the change in cement mix for the HSM concrete to be acceptable based on the small impacts to dose rates as well as that the proposed new cement mixs shielding performance and ability to resist accident conditions does not different from the current cement mix.

6.3 Evaluation Findings

In summary, based upon its review, as described above, the staff has reasonable assurance that the design features relied on for shielding for the Standardized NUHOMS storage system with the 24PTH DSCs, as modified by this amendment, have been adequately identified and evaluated. The evaluation includes appropriate shielding analyses for the configurations that exist during the different stages of storage operations, including the impacts of normal, off-normal, and accident conditions. The evaluation includes dose rate results that are adequate to support evaluation of the Standardized NUHOMSs compliance with the radiation protection requirements in 10 CFR 72.236(d), the occupational doses estimated to result from storage operations using the Standardized NUHOMS with the 24PTH DSCs, and the adequate consideration and incorporation of ALARA principles into the design and operations. The staff reached this finding on the basis of a review that considered the applicable regulations,

29 appropriate regulatory guides, applicable codes and standards, accepted engineering practices, the statements and representations in the SAR, and the staffs confirmatory analyses.

F6.1 The UFSAR, as modified by the proposed amendment, provides specifications of the spent fuel contents to be stored in the Standardized NUHOMS storage system with 24PTH DSCs in sufficient detail that adequately defines the allowed contents and allows evaluation of the shielding design for the proposed contents. The SAR includes analyses that are adequately bounding for the radiation source terms associated with the proposed contents specifications. (10 CFR 72.236(a))

F6.2 The UFSAR, as modified by the proposed amendment, describes the SSCs important to safety that are relied on for shielding in sufficient detail to allow evaluation of their effectiveness for the proposed term of storage. (10 CFR 72.236(b) and 10 CFR 72.236(g))

F6.3 The UFSAR, as modified by the proposed amendment, provides reasonable and appropriate information and analyses, including dose rates, to allow evaluation of the Standardized NUHOMS storage system with the 24PTH DSCs and its compliance with 10 CFR 72.236(d). This evaluation is described in the radiation protection review (chapter 10 of this SER).

F6.4 The UFSAR, as modified by the proposed amendment, provides reasonable and appropriate information and analyses, including dose rates, to allow evaluation of consideration of ALARA in the design of the Standardized NUHOMS storage system with the 24PTH DSCs and evaluation of occupational doses. This evaluation is described in the radiation protection review (chapter 10 of this SER).

7.0 CRITICALITY EVALUATION

The staffs objective in reviewing the applicants criticality evaluation of Amendment No. 18 to the NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel system design is to verify that the spent fuel contents remain subcritical under normal, off-normal, and accident conditions of handling, packaging, transfer, and storage. The applicable regulatory requirements are those in 10 CFR 72.24(c)(3), 72.24(d), 72.124, 72.236(c), and 72.236(g). The staff reviewed the information provided in the amendment request to determine whether the Standardized NUHOMS storage system continues to fulfill the acceptance criteria listed in section 7 of NUREG 2215, Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities.

7.1 Criticality Design Criteria and Features The applicant provided a summary of the proposed changes to the NUHOMS spent fuel storage system in Enclosure 2, DESCRIPTION, JUSTIFICATION, AND EVALUATION OF AMENDMENT 18 CHANGES, to the letter dated May 20, 2022 (ML22140A025). This section of the SER will discuss the proposed changes which affect the criticality safety of the storage system. These proposed changes are as follows:

Change No. 1: Provide for an improved 24PTH basket design using staggered plates similar to EOS-37PTH to simplify construction, reduce weight and improve fabricability.

30 7.2 Proposed Change No. 1 - 24PTH Type 3 Basket Design 7.2.1 Fuel Specification The applicant stated that all the PWR fuel assembly classes acceptable for storage in Type 1 and 2 canisters are allowed for the NUHOMS 24PTH Type 3 canister. TS table 1-1l identifies the fuel specifications for all allowable PWR fuel assembly classes. TS tables 1-1p and 1-1q identify the minimum soluble boron requirements for fuel enriched up to 4.5, 4.6, 4.7, 4.8, 4.9, and 5.0 weight percent 235U for each allowable configuration of intact and damaged/failed fuel.

Table P.6-2 lists the authorized assembly types for the NUHOMS 24PTH canister.

The applicant stated that the NUHOMS 24PTH has three types of canister lengths, each with three allowable configurations on placement of undamaged, damaged, or failed fuel, as shown in figure P.2-6 of the SAR. These configurations include:

Configuration 1: Undamaged fuel in all basket locations Configuration 2: Damaged fuel in up to 12 peripheral basket locations and undamaged fuel in remaining locations Configuration 3: Failed fuel in up to eight peripheral basket locations, damaged fuel in up to four additional peripheral basket locations, and undamaged fuel in remaining locations The applicant used identical minimum soluble boron concentrations for each enrichment for each canister length and dependent on basket rather than canister type because the limitations are derived from a conservative infinite-length system using axial periodic reflection.

7.2.2 Model Specification 7.2.2.1 Configuration The addition of the Type 3 basket changed the storage system response to off-normal and accident conditions. The Type 3 basket introduced a staggered plate structure as opposed to the welded stainless steel fuel tubes sandwiched by either aluminum, poison plates or both aluminum and poison plates. The structural analysis evaluated in SER section 4 adequately demonstrated basket stability. The applicants evaluation of system reactivity excluded deformation. Minor basket distortion occurred under drop conditions, with compressions of the basket as noted in the structural analysis. This reduced the compartment width, a primary distinguisher of the more conservative behavior of the Type 3 basket. SER section 7.2.3.2 provides further details.

The applicant modeled the Type 3 basket using the most reactive configuration of PWR fuel assemblies determined from previous Type 1 and 2 basket analyses. For the basket geometry, the applicant performed sensitivity studies, which are summarized in SAR section P.6.6.4, to determine the most reactive fuel assemblies, enrichments, and soluble boron concentrations.

Examination of nominal and minimum compartment width are demonstrated to be bounded by Type 1 and 2 basket scenarios. This is demonstrated with the most reactive assembly demonstrating reduced reactivity across a range of enrichments and soluble boron concentrations relative to the Type 1 and 2 baskets. The reactivity reduction resulted from increased poison loadings and compartment width. The most reactive configuration consisted of the minimum fuel compartment tube inner dimension, minimum basket structure thickness, minimum assembly-to-assembly pitch, and eccentric positioning of fuel assemblies towards the center of the basket. The applicant also conservatively modeled the MMCs basket walls with a 10B content of 90% of the required minimum specified in table P.6.1 and Drawings NUH24PTH-

31 L-5012-SAR, NUH24PTH-S-5012-SAR, and NUH24PTH-S-LC-5012-SAR. As was done for the Type 1 and Type 2 baskets, the applicant modeled the fuel material as UO2 at a stack density of 97.5% theoretical density with an infinite active fuel length. The applicant also conservatively increased the modeled fuel material and effective height in the assembly by not making allowance for either the dishing or chamfering processes associated with fuel pellet manufacturing. Additionally, the applicant conservatively assumed that the pellet-to-clad gap is filled with the full-density moderator.

For damaged fuel, the applicant modeled the fuel as clad rods with uniform pitch expansion and with UO2 fuel present in the guide and instrumentation tubes. For failed fuel, the applicant continued to model the fuel as intact or partially de-cladded rods with uniform pitch expansion and six inches of poison-uncovered fuel. The applicant modeled select cases representing the most reactive conditions previously determined for the Type 1 and 2 basket analysis at 4.5 and 5.0 weight percent 235U for the Type 3 basket with damaged and failed fuel.

The Type 3 configuration of the NUHOMS 24PTH canister analyzed in the criticality analysis is consistent with previously approved analyses of Type 1 and 2 baskets. The staff finds that the applicant has determined the most reactive configuration of the canister basket for each fuel type.

7.2.2.2 Material Properties The applicant described the fuel and basket materials of the NUHOMS 24PTH in SAR section P.6.3.2 and SAR table P.6-8. These descriptions included the composition of the major components of the NUHOMS 24PTH: UO2 fuel, steel and aluminum structural components, and the MMC neutron absorber. The applicant used fuel materials for the NUHOMS 24PTH Type 3 identical to those previously evaluated and approved for the standard NUHOMS 24PTH Types 1 and 2. The major difference between the NUHOMS 24PTH Type 1 and Type 2 baskets and the NUHOMS 24PTH Type 3 basket is the staggered HSLA composite basket in the Type 3 canister which is identical in material composition and design to that of the previously approved NUHOMS EOS-37PTH canister. The staff investigated the HSLA composition variations to ensure conservatism. The criticality analyses conservatively assumed no more than 90% of the MMC minimum 10B content specified in SAR table P.6.1 and Drawings NUH24PTH-L-5012-SAR, NUH24PTH-S-5012-SAR, and NUH24PTH-S-LC-5012-SAR.

The staff determined that the material properties assumed in the NUHOMS Type 3 basket criticality analyses are consistent with those assumed in previously approved analyses of the NUHOMS Type 1 and 2 and NUHOMS EOS-37PTH basket systems. Therefore, the staff finds that the applicant has determined conservative material properties for the canister basket and each fuel type.

7.2.3 Criticality Analysis Like the previously evaluated Types 1 and 2 baskets, the applicant modeled the B&W 15x15 for the NUHOMS 24PTH Type 3 basket criticality analysis because it is the most reactive PWR fuel assembly type at the examined enrichments. The applicant modeled the fuel and canister basket according to the most reactive combination of fuel and basket parameters determined previously for similar canisters. The applicant used analyses performed for the NUHOMS 24PTH Types 1 and 2 baskets to determine the characteristics of the most reactive system configuration. These analyses included eccentric fuel assembly positioning as well as varying internal and external moderator density. Using the most reactive configuration determined from these analyses with the Type 3 basket model, the applicant determined the maximum system

32 keff for each fuel assembly class with the minimum required soluble boron in TS tables 1-1p and 1-1q.

7.2.3.1 Computer Programs The applicant used the SCALE version 6.0 three-dimensional KENO V.a Monte Carlo neutron transport code and the ENDF/B-V.0 44 group cross-section library for all keff calculations for this amendment. The SCALE code is a standard in the nuclear industry for performing Monte Carlo criticality safety and radiation shielding calculations. The staff performed confirmatory calculations using the CSAS5 sequence of the SCALE 6.2.4 code system, with the KENO V.a three-dimensional Monte Carlo neutron transport program and the continuous-energy ENDF/B-VII.1 cross-section library.

7.2.3.2 Multiplication Factor The applicant demonstrated that keff values for the storage configurations for the NUHOMS 24PTH canister are all below 0.95. The calculated keff values included all biases and uncertainties determined for the canister in the benchmarking analysis for the SCALE version 6.0 code and 44 group ENDF/B-V.0 cross-section library used in the criticality analysis.

Therefore, the NUHOMS 24PTH system with the Type 3 basket will remain subcritical under normal, off-normal, and accident conditions, demonstrating the design of the system is subcritical as required by 10 CFR 72.124(a) and 72.236(c).

Following a request for additional information on detailed engineering drawings and structural analysis, the applicant reduced the minimum compartment width and evaluated the impact on criticality. After performing the revised analysis, the applicant provided a new eigenvalue for the associated deformation. As expected, the minimum compartment width reduction caused an increase in eigenvalue. SAR tables P.6-49a, P.6-50a, and P.6-51a provided the minimum compartment width eigenvalue counterpart to the nominal eigenvalue. The eigenvalue increase proved minor enough to retain conservatism relative to the Type 1 and 2 basket keff. The staff confirmed values that demonstrated the greatest loss in this margin of conservatism.

Additionally, SAR table P.3.8-11 noted compression of the basket due to the side drop test. The applicant failed to analyze compression of the minimum compartment width. The staff investigated the instance with the least conservatism relative to the Type 1 and 2 basket by further reducing the compartment width to account for compression. The staff analysis demonstrated that, while this additional reduction in compartment width increased reactivity, the case with the least initial margin still maintained conservatism relative to the Type 1 and 2 baskets.

The staff performed confirmatory criticality evaluations of the NUHOMS 24PTH system with varying steel compositions in the NUHOMS 24PTH Type 3 basket. Using additional steel compositions for alloy concentrations possible with HSLA steels, the staff calculated keff values for select configurations which were within the margin of error of those calculated by the applicant and confirmed that the storage system is subcritical per the requirements of 10 CFR 72.124(a) and 72.236(c).

7.2.3.3 Benchmark Comparisons The applicant provided a benchmarking analysis of SCALE version 6.0 and the 44 group ENDF/B-V.0 cross-section library for use in evaluating the NUHOMS 24PTH canister. The applicant demonstrated reasonable assurance that the Type 3 basket design is a conservative, less reactive state than the Type 1 and 2 basket designs upon which the benchmarking analysis

33 is based. The applicant detailed the benchmarking analysis in SAR sections P.5 and P.6.5. The applicant also summarized the benchmarking analysis in SAR section P.6.5.3. In addition, SAR table P.6-46 provides trending parameters and benchmark eigenvalues using the most conservative upper subcritical limit. The benchmarking analysis determined that the calculated bias did not exhibit any significant trends. It also demonstrated that the Type 1 and 2 baskets were within the range of applicability of both the bias and the bias uncertainty for all parameters considered. The design of the Type 3 basket involves similar fuel types, basket type, poison, and fuel configurations as the Type 1 and 2 baskets. Therefore, the applicant determined that the previously approved benchmarking analysis performed for the Type 1 and 2 baskets, as well as the code and cross-section data used in this amendment, are still applicable. There are no significant deviations from the type or concentrations of fuel, moderator, and neutron absorber material (the 3D basket neutron absorber loading increased by less than 10% compared to either the 1C or 2C basket) used in previously approved evaluations, and all of the parameters of interest to the criticality calculation remain within the area of applicability of the previous benchmarking analysis. Therefore, the staff finds that the previously approved bias and bias uncertainty for the Type 1 and 2 baskets modeled with SCALE 6.0 and the 44 group ENDF/B-V.0 cross-section library are appropriate for the Type 3 basket criticality calculations for the NUHOMS 24PTH canister.

7.3 Evaluation Findings

The staff concludes that the criticality design features for the NUHOMS 24PTH spent fuel storage system design are in compliance with 10 CFR Part 72 and that the applicable design and acceptance criteria have been satisfied. The evaluation of the criticality design provides reasonable assurance that the Type 3 basket for the NUHOMS 24PTH spent fuel storage system design will allow safe storage of spent fuel. This finding is reached based on a review that considered the applicable regulations, appropriate regulatory guides, applicable codes and standards, and accepted engineering practices.

F7.1 Structures, systems, and components important to criticality safety are described in sufficient detail in UFSAR chapters 3 and 6 to enable an evaluation of their effectiveness. Design dimensions and deformations are described in drawings and findings. Tolerances are not specified in instances, but parameterization of design characteristics in the criticality safety analysis is performed at a reasonable level to develop limiting conditions which are incorporated into maximum reactivity models.

The deformation impact was not evaluated by the applicant, but confirmatory calculations have shown the configuration remains conservative relative to the Type 1 and 2 basket designs.

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

34 F7.4 The 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 CoC application.

8.0 MATERIALS EVALUATION The staff reviewed and evaluated the information provided by the applicant in Amendment No.

18. The specific material changes evaluated in this section include:

Change No. 1: Provide for a 24PTH improved basket design (Type 3) using staggered plates similar to EOS-37PTH to simplify construction, reduce weight and improve fabricability. To implement this change, the applicant increased the weight limit in CoC appendix B TS, table 1-1l, PWR Fuel Specification for the Fuel to be Stored in the NUHOMS - 24PTH DSC for fuel assembly plus control components (CC) from 1682 lbs.

to 1715 lbs. based on the lighter Type 3 basket; deleted the total weight of the FFC and all its contents; deleted similar statements about the total weight of the FFC plus all its contents from table 1-1e for the 32PT DSC and from table 1-1t for the 61BTH DSC; added a new section in CoC appendix A, Inspections, Tests and Evaluations (ITE) to address the high strength low alloy steel used in the new Type 3 basket design for the 24PTH DSC; and updated the transfer cask does rate values for the 24PTH DSC.

Additionally, the applicant revised the CoC, appendices and UFSAR to differentiate 24PTH Type 3 basket requirements from the existing 24PTH Type 1 and Type 2 basket requirements as necessary.

Lastly, the applicant updated chapters 12 and P.1 to reflect ISG-2, Revision 2 with the focus on the new Type 3 basket SSC changes.

Change No. 4: Updated appendix C ASME Code Alternatives Table C-12 to add code alternative NG-4231.1 as approved on February 3, 2022 (ADAMS Accession No. ML22025A169).

Change No. 7: Clarified in appendix B TS LCO 3.1.3 that, consistent with existing UFSAR analysis, no transfer time limit is associated with the 24PTH-S-LC DSC.

Additional Scope Change No. 1: Added an additional scope item to allow the use of blended Portland-Limestone cement certified to the requirements of ASTM C595. The reason for this change is that the cement supplier, as of January 2023, will no longer provide cement in accordance with ASTM C150 as the supplier is transitioning to a cement with a smaller carbon footprint that includes 10% limestone. This change supports continued HSM fabrication activities.

To implement this change, the applicant proposed to change UFSAR sections 4.2, Storage Structures; 4.10, References; P.4.4.8, Evaluation of HSM-H Performance; chapters R.3, Evaluation of HSM Model 152 Concrete Components with Temperature exceeding Code Limits; R.4, Evaluation of HSM Model 152 Concrete Temperatures; V.3, Evaluation of HSM Model 202 Concrete Components with Temperature exceeding Code Limits; and V.4, table V.4-2 Notes in support of this additional scope item.

The staff reviewed the changes to the UFSAR, CoC, and TS associated with the application.

The staffs review included the material incorporated by reference from CoC No. 1004. The staff conducted their review using the guidance in chapter 8 of NUREG-2215, Standard Review Plan

35 for Spent Fuel Dry Storage Systems and Facilities, to conclude there was adequate materials performance under normal, off-normal, and accident-level conditions.

The areas of review covered in this SER section are described in NUREG-2215 Section 8.2, and include system design, engineering drawings, material selection and material properties, environmental conditions and material compatibility, cladding integrity and fuel condition. The staff also evaluated the changes in the application with respect to the 10 CFR Part 72 regulatory requirements identified in NUREG-2215 Section 8.3, and the review procedures and acceptance criteria identified in NUREG-2215 Section 8.4.

In addition to the guidance in NUREG-2215, the staff evaluated the engineered drawings and the description of the SSCs included in the application using the information provided in NUREG/CR-5502, Engineering Drawings for 10 CFR Part 71 Package Approval, and NUREG/CR-6407, Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety.

8.1 Change No. 1 - Lighter 24PTH Type 3 Basket Design The applicant plans to use the following materials to fabricate the Type 3 basket: HSLA, SA-516 Gr70 and ASTM A516 Gr70, ASTM B221 or B209 Alloy 6061-O, and SA-564 Gr. 630 H1100.

The staff finds the proposed Type 3 basket materials acceptable since these structural materials were qualified and tested for the EOS-37PTH under CoC No. 1042 (ADAMS Accession No. ML20136A052). The applicant provided the properties for these materials in UFSAR tables P.3.3-10 through P.3.3-13. The staff reviewed ASME,Section II, Part D, 2010 edition with 2011 addenda. The staff confirmed that the material properties in UFSAR tables P.3.3-11 thru P.3.3-13 are appropriate for the temperature ranges specified in the CoC. The applicant proposed using an ultimate yield strength (Su) that is 1.05 times higher than the ultimate yield strength (Sy), i.e., Su = 1.05 Sy. After reviewing Sy and Su values for SA-517 in ASME Section II Part D, the staff determined that this approach is conservative.

In UFSAR section P.4.2, Summary of Thermal Properties of Materials, the applicant provided emissivity values for the different materials used in the 24PTH Type 3 DSC analyses. After reviewing the literature data, the staff confirmed that emissivity values identified by the applicant are appropriate for use in the thermal analysis; therefore, the staff finds the thermal properties acceptable.

In UFSAR section P.3.4, General Standards for Casks, the applicant proposed an addition describing the materials of the Type 3 basket and its behavior in borated water. The staff determined that the UFSAR section P.3.4 updates were acceptable because the surface treatment of the HSLA steel plates provides adequate corrosion protection during manufacturing and during immersion in borated water. The staff performed a similar evaluation of the surface treatment of HSLA in SER section 8.8.3, High-Strength, Low-Alloy Steel in Deionized Water and Boric Acid for EOS Amendment No. 0 (ADAMS Accession No. ML16242A023). For EOS Amendment No. 0, the staff found the use of high-strength, low-alloy steel in a borated water environment to be acceptable.

In addition, the applicant proposed editorial changes to UFSAR sections P.3.3, Mechanical Properties of Materials, and P.3.4, General Standards for Casks, that differentiate the Type 1 and 2 baskets from the Type 3 basket as well as describe the Type 3 basket material properties.

The staff finds these editorial updates in UFSAR section P.3.3 acceptable because they provide clarity regarding which requirements are associated with the 24PTH Type 1, Type 2 and Type 3 baskets.

36 The applicant also proposed revising the maximum weight limit for a fuel assembly plus control components from 1682 lbs. to 1715 lbs. due to the new Type 3 basket design being lighter than the previous basket designs. As explained in section 4.2 of this SER, the applicant indicated that the weight of the 24PTH DSC is bounded by the weight of the 32PTH1 DSC. Since the staff previously reviewed and accepted the analyses for the 32PTH1 DSC, the staff finds a maximum weight limit of 1715 lbs. for the fuel assembly and control components for the 24PTH DSC acceptable.

The applicant also proposed deleting the total weight of the FFC plus contents for the 32PT, 24PTH and 61BTH DSCs from CoC appendix B tables 1-1e, 1-1l, and 1-1t respectively. The staff determined that this change is consistent with the approach used in CoC Amendment No.

14 (ADAMS Package Accession No. ML18031A106) for the 32PTH1 DSC in which the total weight of the FFC plus contents is not specified. In addition, the staff noted that the FFC plus contents total weight requirement remains specified within the UFSAR to convey the design requirements and to ensure that any modifications to the FFC remain within the applicable bounds of the UFSAR evaluation. Also, the applicant stated in their justification for the change that, if the weight of the FFC exceeds the maximum specified for the FFC in the UFSAR, it is evaluated using the design change and 10 CFR 72.48 process. Therefore, the staff finds deletion of FFC plus contents total weight acceptable.

Due to the basket design changes and the content weight changes discussed above, the applicant made changes to the CoC and its appendices. The applicant added 24PTH and a note with Type 3 basket information to the ASME Code table in CoC Section II.1.b. The staff finds the CoC ASME Code Requirements acceptable because the information identifies the applicability of the 24PTH Type 3 basket requirements. The applicant discussed the HSLA steel used in the Type 3 basket design in the new appendix A section 2.4, High Strength Low Alloy Steel for Type 3 Basket Structure for the 24PTH DSC. The staff previously evaluated the DSC HSLA basket material for EOS Amendment No. 0. In SER section 8.4.2, Dry Shielded Canister Basket Assembly issued with EOS Amendment No. 0, the staff found that the basket materials were consistent with those required by the applicants structural analyses (ADAMS Accession No. ML16242A023). Therefore, the staff finds the basket material requirements in new appendix A section 2.4 are acceptable.

Additionally, the applicant modified the following appendix B tables because of the new basket design and the content weight changes:

tables 1-1l to reference the Type 3 basket and to incorporate the content weight changes discussed above; tables 1-1l, 1-1p, 1-1q, 1-1q1 and 1-1r to reference the Type 3 basket.;

referenced the Type 3 basket in figure 1-15 by adding notes and added the Type 3 basket to the figure 1-15a title.

tables 1-1e and 1-1t to remove the maximum FFC content weight.

The staff finds the CoC appendix B updates to tables 1-1l, 1-1p, 1-1q, 1-1q1 and 1-1r, figures 1-15 and 1-15a acceptable because the applicant clearly differentiated 24PTH Type 3 basket requirements from the existing 24PTH Type 1 and Type 2 basket requirements. The staff finds change to tables 1-1e and 1-1t acceptable because it is consistent with the changes made in appendix B table 1-1l.

The applicant also changed CoC appendix C by updating the List of Tables, by adding a note differentiating 24PTH Type 1 and 2 baskets from Type 3 to the ASME Code Alternatives table and by modifying the table C-8 title to specify 24PTH Type 1 and 2 baskets. The staff finds the

37 appendix C changes acceptable because the applicant clearly differentiated 24PTH Type 3 basket requirements from the existing 24PTH Type 1 and Type 2 basket requirements.

8.2 Change No. 4: NG-4231.1 Code Alternative Addition to Appendix C The applicant updated appendix C ASME Code Alternatives table C-12 to add the recently approved code alternative NG-4231.1. The staff reviewed the additions and found that they are consistent with the ASME code alternative the staff approved for the Standardized NUHOMS 61 BTH Type 2 DSC in CoC No. 1004 Amendment No. 13, Revision 1, and Amendments Nos.

14-17 (ADAMS Accession No. ML22025A169). Therefore, the staff finds this change acceptable.

8.3 Change No. 7: Clarify Appendix B TS LCO 3.1.3 Time Limit The applicant updated the appendix B TS by adding a clarification to the Time Limit for Completion of DSC Transfer for 24PTH-S-LC for the existing Type 2 basket and added a time limit for completion of transfer for 24PTH-S, 24PTH-L and 24PTH-S-LC DSCs with the new Type 3 basket. The applicant also removed DSC from each row as it was redundant to the column heading. The staff reviewed UFSAR section P.4.6.5.2, which describes the 24PTH-S-LC thermal evaluation, and UFSAR table P.4-14 which documents the analytical results. The applicant specified no time limits for the 24PTH-S-LC HLZC 5 which indicates that there is no time limit requirement. The staff finds this change acceptable because it is consistent with the existing UFSAR analysis.

The staff reviewed UFSAR table P.4-54. Table P.4-54 compares the maximum fuel cladding and DSC component temperatures for the 24PTH Type 3 DSC in the HSM-H with HLZC #1 to the design basis values presented in table P.4-14 and table P.4-15 for the normal hot storage condition with a maximum heat load of 40.8 kW and a 100 °F (38 °C) ambient temperature. The staff confirmed that the maximum component temperatures for the 24PTH Type 3 DSC and the HSM-H with HLZC #1 under the normal storage condition in table P.4-54 are bounded by design basis values listed in table P.4-14 and table P.4-15. The staff confirmed that the design basis values in tables P.4-20 and P.4-21 for off-normal storage conditions, as well as tables P.4-25 and P.4-26 for the blocked vent accident condition, also remain bounding for the 24PTH Type 3 DSC. Therefore, the staff finds that the fuel cladding and component temperatures will not exceed the maximum allowed temperatures.

The staff reviewed the remaining additions for the Type 3 basket, 24PTH-S and 24PTH-L. The staff finds them acceptable because they are consistent with the existing thermal analyses. The staff also reviewed the removal of DSC and agreed that it was redundant; therefore, the staff finds the change acceptable.

8.4 Additional Scope Change No. 1: Change Regarding Blended Portland-Limestone Cement The applicant added an additional scope item to allow the use of blended Portland-Limestone cement that would be certified to the requirements of American Society of Testing and Materials (ASTM) C595, Standard Specification for Blended Hydraulic Cements. The applicant explained that, as of January 2023, their cement supplier had transitioned to a cement with a smaller carbon footprint that includes 10% limestone. As a result, the cement supplier no longer provides cement in accordance with ASTM C150, Standard Specification for Portland Cement, which allows up to 5% limestone.

38 This resulted in UFSAR changes to the following portions of the UFSAR:

Section 4.2, Storage Structures Section 4.10, References Section P.4.4.8, Evaluation of HSM-H Performance Chapter R.3, Evaluation of HSM Model 152 Concrete Components with Temperature exceeding Code Limits Chapter R.4, Evaluation of HSM Model 152 Concrete Temperatures Chapter V.3, Evaluation of HSM Model 202 Concrete Components with Temperature exceeding Code Limits Chapter V.4, table V.4-2, Notes In UFSAR section 4.2, Storage Structures, the applicant added the option to use blended Portland cement. In UFSAR section P.4.4.8, Evaluation of HSM-H Performance, the applicant also added a condition that HSM-H model concrete testing is performed when the concrete accident temperature exceeds 350 °F (177 °C). The applicant implemented this requirement to ensure that both ordinary Portland cement (OPC) meeting ASTM C150 Type II and blended hydraulic cement, e.g., Portland-Limestone cement (PLC), meeting the specifications of ASTM C595 retain the required strength and remain acceptable for use. The applicant required testing when the concrete accident temperature exceeds 350 °F (177 °C) to demonstrate that the level of strength reduction is less than the 10% penalty that was employed in the calculations and to ensure that concrete deterioration due to the elevated temperatures does not occur.

In addition, the applicant added text to appendix A, section 4.2 for testing HSM-H storage modules. The staff noted that the CoC No. 1004 appendix A, 4.2 requires testing of compressive strength for elevated temperatures whenever there is a significant change in the cement, aggregates, the cement supplier, or water-cement ratio of the concrete mix design. The CoC No. 1004 appendix A, 4.2 changes remain consistent with NUREG-2215 Section 8.5.8.2.

Concrete Design and Temperature Limits, that includes specific guidance for concrete components operated above the ACI 349 appendix E temperature limits. The staff noted that ACI 349 allows multiple ASTM specifications for cement including both ASTM C150 and ASTM C595. Therefore, the staff determined that the concrete testing stipulated in the applicants TS is adequate to verify the performance of concrete using either ASTM C150 or ASTM C595 cement at elevated temperatures.

The staff independently evaluated the UFSAR and Appendix A changes proposed by the applicant by reviewing three reports. The first report, CALTRANS: Impact of the Use of Portland-Limestone Cement on Concrete Performance as Plain or Reinforced Material, dated June 29, 2021, aimed to address whether PLC could replace OPC without loss of mechanical and durability performance. The report evaluated alkali-silica reactivity (ASR), shrinkage and restrained shrinkage cracking, mechanical properties, transport properties, chloride binding, resistance to chloride ingress in concrete, corrosion of reinforcing steel, air entrainment, and external sulfate attack. The report concluded that, when compared to OPC, PLC had similar or improved ASR performance, statistically similar shrinkage, flexural strength, set times, bound chloride contents, comparable porosity, formation factor and chloride apparent diffusion coefficient, similar critical chloride thresholds and time to corrosion initiation, similar or slightly improved performance when exposed to sulfate. Additional benefits of PLC included the potential for a 10-12% reduction of greenhouse gas emissions. Overall, the report concluded that PLCs could directly substitute OPC one for one in concrete mixtures.

The second report, Georgia Department of Transportation Assessment of Limestone Blended Cements for Transportation Applications, dated September 2017 compared Type I/II and Type IL (Portland-Limestone blended) cements from five producers to determine material

39 characteristics such as setting time, strength development, shrinkage, and permeability and mechanical properties. The assessment identified that the performance of concrete made with Type IL cement was affected by the cement fineness which increased strength and drying shrinkage. The report concluded that Type IL cement may be used in place of Type I/II cement when fineness values are specified. Chapter 7 of the report described the durability testing. The description included rapid chloride permeability test, surface resistivity test, and freezing and thawing resistance. The results of the rapid chloride permeability test indicated little difference between the electrical properties for the Type I/II and Type IL cements. The report also compared Type I/II and Type IL mixtures and identified no consistent trends in surface resistivity based on either limestone dosage or fineness. The freezing and thawing results showed the Type IL cement concrete freeze and thaw resistance and Type I/II cement concrete to be similar.

The third report, State-of-the-Art Report on Use of Limestone in Cements at Levels of up to 15% by the Portland Cement Association (PCA), dated September 2014 assessed mechanical properties: strength, strength development and volume stability. The report also evaluated durability characteristics such as permeability, chloride resistance, carbonation, freeze thaw, deicer salt scaling, sulfate resistance, alkali-silica reaction, and abrasion resistance. With regards to compressive strength, tensile strength, flexural strength and modulus of elasticity, the study concluded that there was no significant difference between Portland Cement (PC) and PLC. Tests evaluating permeability, chloride resistance, carbonation, freeze/thaw and deicer salt scaling, sulfate resistance, alkali-silica reaction, and abrasion resistance performance showed no significant difference between PC and PLC concrete.

After reviewing the three independent reports, the staff concluded that the mechanical performance of PLC is either comparable to or better than OPC. The staff also concluded that the durability of PLC is either comparable to or better than OPC. Therefore, the staff finds the use of PLC acceptable.

In UFSAR section 4.10, References, the applicant added ASTM C595. The staff noted that, because the requested change did not specify a specific section of ASTM C595, the requested change could allow the use of multiple types of PLC which may not have similar properties.

Therefore, the staff reviewed available information on the effects of blended hydraulic cement containing Pozzolans1 as specified in ASTM C595. Specifically, the staff reviewed the information in American Concrete Institute (ACI) 232.1R-18, Report on the Use of Raw or Processed Natural Pozzolans in Concrete, and ACI 232.2R, Report on the Use of Fly Ash in Concrete. These reports indicated that concrete containing a pozzolan typically provides lower permeability; reduced heat of hydration; reduced alkali-aggregate reactivity; higher strengths at later ages; and increased resistance to attack from sulfates, compared with concrete that does not contain pozzolan. Fly ash is used in concrete and other Portland cement-based systems primarily because of its pozzolanic and cementitious properties. The partial substitution of fly ash for Portland cement has been shown to significantly reduce chloride permeability.

The staff also reviewed available information on the effects of blended hydraulic cement containing slag as specified in ASTM C595. Specifically, the staff reviewed the information in 1 As defined in ACI 116R-00, Cement and Concrete Terminology, a Pozzolan is a siliceous or siliceous and aluminous material that in itself possesses little or no cementitious value but that will, in finely divided form and in the presence of moisture, chemically react with calcium hydroxide at ordinary temperatures to form compounds having cementitious properties. The name is derived from Pozzuoli, which is now part of Naples, Italy, where the Romans mined volcanic sand that was used in making concrete.

40 ACI 226.1R-87, Ground Granulated Blast-Furnace Slag as a Cementitious Constituent in Concrete, and ACI 233R-17, Guide to the Use of Slag Cement in Concrete and Mortar. These reports showed that the addition of slag increases strength after curing, improves sulfate resistance and reduces permeability which reduces chloride migration without affecting the modulus of elasticity, creep and shrinkage, freeze-thaw resistance, or the passivity of the reinforcing steel.

The staff also reviewed available information on the ternary cement blends2 that are included in ASTM C595 in the testing reports by PCA, GDOT, Caltrans and the guidance in ACI 233R-17.

These reports generally indicated that the durability of concrete can be increased using ternary cement blends. Testing results documented in these reports showed that ternary blend cement has equivalent early-age strength and improved later-age strength compared to Portland cement. Ternary blends also have improved chloride and sulfate resistance, as well as increased resistance to ASR.

The staff reviewed available information on concrete temperature stability summarized in NUREG/CR-7031, A Compilation of Elevated Temperature Concrete Material Property Data and Information for Use in Assessments of Nuclear Power Plant Reinforced Concrete Structures. The conclusions in NUREG/CR-7031 stated that compressive strength at elevated temperatures decreased as the cement content increased. Similarly, concretes with lower cement content experienced less tensile strength reduction than those with higher cement content. Also, concrete containing fly ash exhibited a strength increase in the temperature range of 121 to 149 °C (250 to 300 °F).

NUREG/CR-7031 concluded that the aggregate-cement paste bond region has been shown to be the weakest link because it is normally weaker than the cement paste which is normally weaker than the aggregate. Test results documented in NUREG/CR-7031 showed that a significant strength reduction for concrete was observed when the aggregate-cement paste bond failed because of increased temperatures, chemical interactions or thermal incompatibility between the aggregate and cement paste even if both the aggregate and surrounding mortar matrix remained intact.

Based on the information reviewed, the staff determined that concrete produced using ASTM C595 blended cements will retain adequate strength and durability when the guidance in NUREG-2215 Section 8.5.8.2 is followed.

8.5 Management of Aging Mechanisms and Effects The applicant performed an aging management evaluation in the amendment application. The application introduced a new basket design and materials for the 24PTH DSC. In addition, the application included a change to allow the use of a blended Portland-Limestone cement that would be certified to the requirements of ASTM C595. The applicant evaluated the impact of these changes on the NUHOMS Aging Management Plan. The applicant determined that there were neither new aging effects requiring management nor new aging management activities required for any of the added subcomponents. The applicant concluded that (1) no Amendment No. 18 UFSAR changes caused any additions to the Standardized NUHOMS System CoC No.

1004 renewal UFSAR changes, and (2) no Amendment No. 18 technical specification changes 2 As defined in ASTM C595, Standard Specification for Blended Hydraulic Cements, a ternary cement blend is blended hydraulic cement consisting of Portland cement with either a combination of two different pozzolans, slag and a pozzolan, a pozzolan and a limestone, or a slag and a limestone.

41 caused additions to the Standardized NUHOMS System CoC No. 1004 renewal technical specification changes.

The staff reviewed the application and determined that the applicant did not propose any changes that affect the staffs materials evaluation provided in previous safety evaluations for CoC No. 1004, Renewed Amendments Nos. 1 through 17. The staff determined that the applicants potential aging mechanisms assessment of the new 24PTH DSC basket materials and design is consistent with the assessment in NUREG-2214, Managing Aging Processes in Storage Report, for the material and environment combination. The staff also finds acceptable the applicants conclusions that there are no credible aging mechanisms. For the proposed change to allow the use of a blended Portland-Limestone cement, the staff determined that change added no unique aging effects requiring a UFSAR/Technical Specifications change because the potential aging mechanisms and effects for blended cement concrete are the same as those for Portland cement concrete. Based on its review, the staff finds that the applicant has considered the materials and specific environments to identify the aging mechanisms that could lead to loss of intended functions in a manner consistent with NUREG 1927, in its application for the renewal of the Standardized NUHOMS System and therefore, the staff finds the existing time-limiting aging analyses and aging management program to be acceptable.

8.6 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 SNF.

F8.3 The applicant has met the requirements in 10 CFR 72.236(h). The materials of the SNF 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). SNF specifications have been provided.

9.0 CONFINEMENT EVALUATION The staff reviewed and evaluated the information provided by the applicant in Amendment No.

18. The specific confinement changes evaluated in this section include:

Change No. 1: Provide for an improved 24PTH basket design (Type 3) using staggered plates similar to EOS-37PTH to simplify construction, reduce weight and improve fabricability.

The applicant proposed a newly designed NUHOMS 24PTH Type 3 DSC to store intact, damaged, and failed PWR fuel assemblies. The 24PTH Type 3 DSC basket structure consists of composite plates of steel, neutron poison, and aluminum plates that fit together to form the grid structure. The confinement review is to ensure the newly designed NUHOMS 24PTH Type

42 3 DSC has no changes in confinement system and boundary and the various features of the new basket design will not degrade confinement performance.

9.1 Confinement System Evaluation The applicant designed the 24PTH Type 3 DSC as a welded vessel to provide confinement of fuel assemblies in an inert atmosphere. The primary confinement boundary consists of the DSC shell, the top and bottom inner cover plates, the siphon and vent block, the siphon and vent port cover plates, and the associated welds, as shown in UFSAR figures P.3.1-1 and P.3.1-2. The outer top cover plate and associated welds form a redundant confinement boundary.

The 24PTH Type 3 DSC confinement boundary remains unchanged from those of the 24PTH Type 1 DSC (with aluminum inserts) and Type 2 DSC (without aluminum inserts). The applicant designed the 24PTH Type 3 DSC confinement boundary to be leaktight with an acceptance criterion of 10-7 ref-cm3/sec, as described by ANSI N14.5. The 24PTH Type 3 DSC lid-to-shell weld met the guidance of ISG-18, which was incorporated into NUREG-2215, such that the leakage of radiological matter from the confinement boundary is non-credible. Therefore, the applicant did not perform a confinement dose analysis. The applicant committed to verify the confinement function of the DSCs, including 24PTH Type 3 DSC, through pressure testing and helium leak testing.

The staff reviewed the UFSAR 1004 Amendment 18 and the Proposed Amendment 18 Changes. The staff determined that the 24PTH Type 3 DSC lid-to-shell weld meets the guidance of ISG 18 which was incorporated into NUREG-2215. Staff also determined that there is no change in the 24PTH Type 3 DSC confinement design, as utilized for 24PTH Type 1 and Type 2 DSCs that were previously reviewed and approved by the NRC. Therefore, the staff finds that any radiological releases to the environment from the 24PTH Type 3 DSC are still within the limits established by the regulations in 10 CFR Part 72 (10 CFR 72.104 and 72.106).

The staff also reviewed UFSAR section P.4.12.1.5.4, Maximum Internal Pressures, and confirmed the maximum internal pressures in UFSAR table P.4-19, table P.4-24, and table P.4-29 remain bounding and are below the design limits for the 24PTH Type 3 DSC under normal, off-normal, and accident storage conditions, respectively.

The applicant did not request changes to the confinement criteria for SSCs important to safety.

For this reason, the staff concludes the confinement criteria continue to comply with the general criteria established in 10 CFR Part 72.

9.2 Evaluation Findings

F9.1 FSAR section P.3.1 describes NUHOMS 24PTH Type 3 DSC confinement structures, systems, and components important to safety in sufficient detail to permit evaluation of their effectiveness.

F9.2 The design of the NUHOMS 24PTH Type 3 DSC adequately protects the spent fuel cladding against degradation that might otherwise lead to gross ruptures. SER chapter 4, Thermal Evaluation discusses the relevant temperature considerations.

F9.3 The design of the NUHOMS 24PTH Type 3 DSC provides redundant sealing of the confinement system closure joints by using dual welds on the canister lid and closure as required by 10 CFR 72.236(e).

43 F9.4 The NUHOMS 24PTH Type 3 DSCs have no bolted closures or mechanical seals.

The confinement boundary contains no external penetrations for pressure monitoring or overpressure protection. Since the NUHOMS 24PTH Type 3 DSC uses an entirely welded redundant closure system, no direct monitoring of the closure is required.

F9.5 The confinement system of NUHOMS 24PTH Type 3 DSC is leaktight for normal conditions and anticipated occurrences, thus the confinement system will reasonably maintain confinement of radioactive material. SER chapter 10, Radiation Protection Evaluation shows that the NUHOMS 24PTH Type 3 DSCs satisfy the regulatory requirements of 10 CFR 72.104(a) and 10 CFR 72.106(b).

F9.6 The confinement system of NUHOMS 24PTH Type 3 DSC has been evaluated by analysis. Based on successful completion of specified leakage tests and examination procedures, the staff concludes that the confinement system will reasonably maintain confinement of radioactive material under normal, off-normal, and credible accident conditions.

F9.7 The staff concludes that the design of the confinement systems of the NUHOMS 24PTH Type 3 DSC complies with 10 CFR Part 72 and that the applicable design and acceptance criteria have been satisfied. The evaluation of the confinement system design provides reasonable assurance that the NUHOMS 24PTH Type 3 DSC will allow safe storage of spent fuel. This finding is reached on the basis of a review that considered the applicable regulations, appropriate regulatory guides, applicable codes and standards, the applicants analyses, and acceptable engineering practices.

10.0 RADIATION PROTECTION EVALUATION The purpose of the radiation protection review is to determine if the storage system, with the proposed changes, complies with the regulatory requirements for radiation protection and to ensure that its design and operations include reasonable consideration of, and facilitate licensees compliance with, the requirements that licensees using the system must meet. This includes evaluation of the compliance with 10 CFR 72.236(d) and appropriate consideration of ALARA among others.

For this amendment, the applicant proposed the following changes that are relevant to radiation protection:

Change No.1: a modified basket design for the 24PTH DSC variants. As part of this change, the applicant increased the maximum allowed weight for the contents (fuel assembly and control components). As part of this change, the applicant modified the lead in both shield plugs of the 24PTH-S-LC DSC to allow for the use of precast lead as an option to the poured lead in the currently approved design. The modification includes differences in lead density and thickness and the introduction of radial gaps between the lead and the steel components of the shield plugs. For this change there were also associated changes to the dose rate limits in sections 3.2 and 3.3.2 of appendix A to the CoC. The staff also identified that other differences were proposed for application of dose rate limits to the different 24PTH DSC variants in the TC and the HSM.

Change No. 3: modifications to the language related to TCs with liquid neutron shields (NS) regarding the OS197L TC in section 4.3.2 of appendix B to the CoC for the purpose of clarification.

44 Change No. 5: removal of the reference to 10 CFR Part 20 from the text in section 4.3.2 of appendix B to the CoC.

Change No. 6: modifications to section 4.3.2 of appendix B to the CoC for the purpose of clarifying the language regarding draining the annulus between the TC and the DSC.

Change No. 9: to incorporate administrative controls during short duration operations that are not analyzed for tornado hazards is also relevant. The aim of instituting the administrative controls is to ensure against an accident resulting from a tornado hazard that puts the storage system in a configuration that challenges the bases for determining the system meets the radiation protection requirements in 10 CFR Part 72 which include not exceeding the regulatory dose limits in 10 CFR Part 72. The adequacy of these administrative controls for short duration operations is evaluated in SER section 11. The adequacy of these controls is a basis for the staffs finding of adequacy of the radiation protection analyses.

Additional Scope Change No. 1: Lastly, the applicant requested an additional change for the HSM concrete, to allow use of different cement.

10.1 Modified 24PTH Basket and 24PTH-S-LC DSC Shield Plug Lead The addition of the new Type 3 basket for all the 24PTH DSCs and the modification of the DSC lead shield plugs for the 24PTH-S-LC had the potential to impact the public dose assessments and compliance with 10 CFR 72.236(d) which points back to 10 CFR 72.104 and 10 CFR 72.106. These changes also had the potential to affect the occupational dose assessments and the implementation of ALARA or the 24PTH DSCs.

The applicant did not provide any new assessments or revise the existing assessments for public dose and compliance with 10 CFR 72.236(d). The applicant explained its analysis methods for calculating doses for 10 CFR 72.236(d) assessments in response to staffs questions on the subject. For normal operations and anticipated occurrences, i.e., off-normal conditions, the applicant described how the calculations rely on the HSM average surface dose rates. Since the applicants shielding analysis indicated that the current average surface dose rates either dont change or are bounding for the 24PTH DSCs with the proposed changes, the changes did not impact the normal operations and anticipated occurrences dose assessments; they either remain applicable to or bounding for the modified 24PTH DSCs. SER section 6.1.4.3 describes the staffs review of the HSM average surface dose rates. Based on the outcome of that review, the staffs evaluation indicated there could, in some cases, be a slight increase in the annual doses estimated for the various evaluated distances. However, the staff also determined that any change in the distance required to avoid exceeding the 10 CFR 72.104(a) dose limits would be minimal. For the remaining instances, the current analysis remains bounding for those cases because the average surface dose rates would decrease which means that the 10 CFR 72.104(a) dose limits would not be exceeded at shorter distances.

Based on this evaluation and the minimal impact for any instance where dose rates might increase due to the changes, the staff finds the applicants current analysis for 10 CFR 72.104 compliance continues to be adequate and acceptable.

For the case of accident doses to the public, the staff identified that the applicants current analysis indicates significant margin to the limits in 10 CFR 72.106(b). Thus, the staff determined that any increases in accident doses due to the changes to the 24PTH DSCs would still be significantly below the dose limits. The staff reviewed the accident analyses for the NUHOMS system with the 24PTH DSCs. The TC accidents primarily resulted in loss of the TC radial neutron shielding. Since radial dose rates decreased with the new Type 3 basket, the staff

45 finds that the current TC accident analysis remains bounding for the 24PTH DSCs in the TC. As noted in SER section 6.1.3.1, the staffs review indicated the possibility of significant dose rates at the TC axial ends from damaged fuel reconfiguration and that even the TC normal conditions axial dose rates with the 24PTH-S-LC exceeded the accident radial dose rates. However, the staff determined that the overall largest accident dose rates are still the TC radial dose rates for the 24PTH-L; thus, the staff finds that evaluating the TC accident conditions based only on the radial dose rates is adequate for this amendment.

For the HSMs, the accident that had any potentially notable effect on doses occurred with HSM models that require a separation distance between them to allow airflow because the inlet and outlet vents are on the sides of the HSM. The applicant anticipated that seismically induced movement could increase the distance between modules. Since those dose rates are dominated by the radiation levels on the side of the DSCs, which decrease with the new Type 3 basket, the staff finds that the current accident analysis for the HSMs with the 24PTH DSCs remains bounding. Based on this evaluation, the staff finds the current accident analysis is bounding for public doses for the proposed changes to the 24PTH DSCs.

The applicant noted in the application that the changes would affect occupational dose estimates for system operations. However, the applicant predicted an overall operations dose estimate increase of only approximately three percent. The staff anticipated the increase could be greater than that due to the increased maximum dose rates expected for the TC axial top in the different operations configurations as well as the number of operations involving personnel performing tasks either on or around the top of the transfer cask. The applicant provided additional information regarding the occupational dose estimates. This information included distances relative to the different TC and HSM surfaces where personnel would be expected when performing the operations and the average dose rates for those areas. The applicant justified the use of average dose rates based on the implementation of ALARA practices which would ensure that personnel minimize their time in any high dose rate location. The applicant also provided a comparison of the average dose rates for these areas for a 24PTH-L DSC with a current approved Type 2 basket and the new Type 3 basket. The staff evaluated the differences in the average dose rates for these two baskets and found they align fairly well with the staffs own dose rate change estimates as discussed in SER section 6.1.4.3. The staff considered the applicants approach, which considers good ALARA practices, for estimating the occupational doses and finds it reasonable. Based on the reasonableness of the approach and the similarity in dose rate differences with the staffs expectations, the staff finds the applicants occupational dose impact evaluations acceptable.

The staffs review and findings regarding the modification to the dose rate limits in appendix A to the CoC for the 24PTH DSCs are described in SER section 6.1.5.

10.2 Changes to CoC Appendix B Section 4.3.2 The applicant proposed various changes to section 4.3.2, Radiation Protection Program, of appendix B to the CoC. The applicant explained that one change reformatted the section content to distinguish which requirements apply to the OS197L TC and which requirements apply to other TCs with liquid NSs. After reviewing this change, the staff finds it acceptable because the change, as revised to ensure language consistency across common requirements, in general preserves the requirements applicable to the relevant transfer casks.

The applicant also proposed to revise the monitoring requirements for operations associated with draining of the annulus between the TC and the DSC and draining of the DSC cavity to prevent inadvertently draining the liquid NS. The applicant introduced these requirements, which

46 include verifying the transfer casks liquid NS is filled and monitoring it during any draining operations of the DSC cavity and the annulus between the DSC and the TC, in Amendment No.

11 for the NUHOMS system (see section 6.5 of the SER issued with Amendment No. 11, Revision 0, in ADAMS Accession No. ML14010A486). The applicant added these requirements because of an instance where the NS was inadvertently drained during cask operations. In justifying the proposed change in the current amendment request, the applicant only stated that the DSC cavity drain is at the opposite end of the DSC from the annulus and NS drains. After reviewing the TC drawings and the Amendment No. 11, Revision 0, SER, the staff failed to see both the validity and sufficiency of the applicants justification. The drawings appear to indicate valves on the NS at the same end of the DSC as the DSC cavity drain. Also, the applicants justification did not appear to address the staffs review that led to this requirement as documented in the Amendment No. 11, Revision 0, SER. Therefore, the staff determined that there should be other means, e.g., operating procedures, in addition to the location differences described by the applicant in the current request to ensure against accidental draining of the NS during DSC cavity draining operations. These means should be sufficiently described in the UFSAR, with justification provided as to why these means are sufficient to ensure that inadvertent draining of the NS will not occur even with the proposed technical specification change. After further review of the CoC appendices and UFSAR descriptions for the relevant operations and considering the importance of the neutron shielding to radiation safety and the shielding it provides, the applicant withdrew the proposed change and restored the affected language.

The applicant also proposed to remove the requirement that the licensee, as part of its 10 CFR 72.212 evaluation, perform an analysis to confirm the 10 CFR Part 20 limits will be satisfied.

The applicant stated that this is because the purpose of the 72.212 evaluation is to confirm compliance with 10 CFR 72.104 limits. While that is the purpose in the regulation, the CoC and its appendices can specify that the evaluation also address other items as needed. The staff added the requirement to confirm the 10 CFR Part 20 limits will be satisfied to the NUHOMS system technical specifications in Amendment No. 11. In Amendment No. 11, the applicant added the OS197L transfer cask, for which there were significant shielding and radiation protection concerns, as described and evaluated in the Amendment No. 11, Revision 0, SER.

The applicants request to remove the requirement indicates that some of the significance and purpose of that requirement, i.e., confirming limits of Part 20 will be satisfied as part of the 10 CFR 72.212 evaluation, is no longer clear because of the applicants reformatting section 4.3.2 for the graded approach amendment, i.e., Amendment No. 16. Given the concerns with the transfer casks in the NUHOMS system, particularly the OS197L, that led to the inclusion of this and other requirements, the staff determined additional justification that addresses the relevant points from the staffs review of Amendment No. 11, Revision 0 would be needed for removing this requirement. After considering that information, the applicant withdrew the proposed change and restored the affected language.

Based on the foregoing considerations and that the remaining changes in section 4.3.2 of appendix B to the CoC are limited to formatting changes that preserve and correctly indicate, or describe, the requirements applicable to each TC type as described in the same section for the Amendment No. 17 CoC, the staff finds the changes to be acceptable.

10.3 HSM Concrete Cement Mixture Change Based on the staffs review and findings described in SER section 6.3, the staff finds that the cement mixture change will minimally impact the radiation protection design of the system and the occupational dose and public dose estimates, including the implementation of ALARA in design and operations of the system. Thus, the staff finds the change acceptable.

47 10.4 Evaluation Findings Based on its review, as described above, the staff finds that the radiation protection design of the Standardized NUHOMS, with the 24PTH DSCs as modified in the amendment, is in compliance with 10 CFR Part 72 and that the applicable design and acceptance criteria have been satisfied. The evaluation of the radiation protection design provides reasonable assurance that the Standardized NUHOMS with the modified 24PTH DSCs will allow safe storage of the allowed spent fuel and control component contents for the 24PTH DSCs. The staff reached this finding based on a review that considered applicable regulations and regulatory guides, codes and standards, accepted health physics practices, statements and representations contained in the SAR, and the staffs confirmatory analyses.

F10.1 The Standardized NUHOMS storage system with the 24PTH DSCs provides radiation shielding and confinement features that are sufficient to meet the requirements of 10 CFR 72.104 and 10 CFR 72.106, in accordance with 10 CFR 72.236(d).

F10.2 The design and operating procedures of the Standardized NUHOMS storage system with the 24PTH DSCs provide acceptable means for controlling and limiting occupational radiation exposures within the limits given in 10 CFR Part 20 and for meeting the ALARA objective with respect to exposures, consistent with 10 CFR 20.1101(b).

F10.3 The Standardized NUHOMS storage system with the 24PTH DSCs includes, to the extent practical and appropriate, adequate features, operating procedures, and controls that are designed to assist a general licensee to meet the radiological protection criteria in 10 CFR 72.126(a) and 10 CFR 72.126(d).

11.0 OPERATING PROCEDURES The operating procedures review ensures that the application presents acceptable operating sequences, guidance, and generic procedures for key operations. The review also ensures that the application incorporates and is compatible with the applicable operating control limits in the technical specifications. The applicant requested this Amendment No. 18 scope item to address Action 3(b) of Enforcement Guidance Memorandum 22-001, Enforcement Discretion for Noncompliance of Tornado Hazards Protection Requirements at Independent Spent Fuel Storage Installations (ISFSIs).

For this amendment, the applicant proposed the following changes that are relevant to operating procedures:

Change No. 9: Incorporate administrative controls during short duration ISFSI handling operations that are unanalyzed for tornado hazards in accordance with the guidance contained in NRC EGM 22-001, "Enforcement Discretion for Noncompliance of Tornado Hazards Protection Requirements at Independent Spent Fuel Storage Installations".

11.1 Change No. 9: Short Term Operational Controls The applicant added text to UFSAR section 3.2.1, Tornado and Wind Loadings, stating that implementation of compensatory measures via administrative controls for short term operations which are not analyzed for tornado hazards is required during ISFSI handling operations. In UFSAR section, 5.1.1.5, Transfer Cask Downending and Transfer to ISFSI, the applicant

48 identified acceptable compensatory measures for general licensees to implement via administrative controls during Standardized NUHOMS System short-term cask transfer operations. The applicant modified UFSAR section 5.1.1.8, DSC Retrieval from the HSM, as well as sections 8.1.5, Transfer Cask Downending and Transfer to ISFSI, and 8.2.1, DSC Retrieval from the HSM, in appendices K, M, P, T, U, Y and Z by adding a Note directing GLs to ensure the administrative controls in UFSAR section 5.1.1.5 are implemented. Since performance of aging management activities may require placing SSCs into configurations not analyzed for tornado hazards, the applicant also revised UFSAR section 12.3, Aging Management Program, directing that the compensatory measures be applied to aging management activities. The staff finds that modification of UFSAR section 3.2.1 is appropriate because it explains how the administrative controls mitigate the effects of tornadoes. Staff also finds that modification of the remaining UFSAR sections is appropriate because these sections cause the unanalyzed conditions. In addition, the staff finds that administrative controls to restrict short term cask transfer operations during projected periods of adverse weather to mitigate tornado hazards form an appropriate basis for demonstrating compliance with 72.236(l) in lieu of engineered controls because implementing engineering controls would not be practical.

The applicant directs general licensees to develop, revise, or review existing procedures for the purpose of implementing compensatory measures by establishing administrative controls that mitigate tornado hazards during handling operations. The administrative controls serve to protect important to safety ISFSI SSCs placed into configurations not analyzed for tornado hazards during short-term operations, e.g., during the time between removing the lid of a loaded TC and installing the HSM door when inserting the DSC into a HSM; during the time a HSM door is removed for aging management activities; and during the time a TC is rotated from a vertical to a horizontal orientation outdoors. The applicant limited the acceptance criteria for beginning operations to ensuring that tornado watches, advisories, or warnings were not expected during the duration of handling operations. In addition, appropriate compensatory actions identified by the applicant include:

1. minimizing, to the extent practicable, the duration of ISFSI handling operations that place ISFSI SSCs that are important to safety in an unanalyzed condition,

2. identifying and securing any potential hazards that could hamper short-term operations during periods of adverse weather or during periods when adverse weather is predicted via performance of site walkdowns,

3. determining the expected bounding duration of the handling operation by either benchmarking or dry runs,

4. identifying the forecasted weather conditions for a time period that both bounds the expected duration of the handling operation and provides a contingency margin of time in the event completion of the handling operation is delayed,

5. ensuring that there are no tornado watches, advisories, or warnings within the expected handling operation duration using one of the following sources:

o the National Weather Service,

o the National Oceanic Atmospheric Administration, Weather Forecast Office nearest the site, or o another source which can be justified as providing equivalent information in terms of timeliness and accuracy,

6. prescribing specific times and frequencies to check the weather forecast, which are not greater than the expected bounding duration of the handling operation, that account for the following factors:

o the area(s) in which short term operations will occur, o the site configuration, and

49 o the time necessary to bring the system into an analyzed configuration,

7. assigning staff to monitor weather during handling operations, and

8. placing systems in a safe and analyzed condition as soon as practicable any time the above conditions cannot be met during handling operations.

The applicant also directs that general licensees document weather checks prior to starting and during handling operations using a log or checklist, record satisfactory completion of these criteria and maintain these records with the dry storage cask campaign documentation.

The staff finds that the administrative controls proposed by the applicant ensure that Standardized NUHOMS system cask transfer activities can occur safely for the following reasons. First, the proposed administrative controls preclude short term handling operations during periods of actual adverse weather events or when adverse weather is predicted to occur.

Second, the compensatory measures protect SSCs that are important to safety by either neutralizing dangers to the SSCs before beginning handling operations, or placing them in configurations that conform to approved engineering analyses after short term operations commence. Third, the staff determined that the NWS, which is an agency under the National Oceanic and Atmospheric Administration, within the Department of Commerce is a reliable weather information source. National Weather Service weather forecasting is recognized as being accurate and reliable in the time periods associated with short duration transfer operations. This finding on the use of administrative controls applies solely to wind and tornado accidents. The staff also notes that these administrative controls will not ensure against a DSC ancillary system, e.g., a crane or other handling system, being placed into a condition which exceeds its design basis loading due to normal or off-normal winds during handling operations.

Accordingly, general licensees need to perform additional evaluations and identify wind gust acceptance criteria that identify appropriate wind speeds for operation of SSCs. These evaluations and wind gust acceptance criteria shall be stipulated and addressed in the related design and licensing basis documents.

11.2 Evaluation Findings The staff concludes that the operating procedures for the NUHOMS standardized system are in compliance with 10 CFR 72.236(l) in that the spent fuel storage cask and its systems that are important to safety have been demonstrated to reasonably maintain confinement of radioactive material under wind and tornado accident conditions during short term operations, and that applicable design and acceptance criteria have been satisfied. The evaluation of the description of operations provides reasonable assurance that the Standardized NUHOMS system will allow safe storage of spent fuel. This finding is reached based on a review that considered the applicable regulations, and accepted engineering practices. Some of the key findings from the staffs review of Amendment No. 18 include:

F11.1 The Standardized NUHOMS System is compatible with dry loading and unloading in compliance with 10 CFR 72.236(h). General procedure descriptions for these operations are summarized in chapter 5 of the applicants SAR. Detailed procedures will need to be developed and evaluated on a site-specific basis.

12.0 CONDUCT OF OPERATIONS For this amendment, the applicant proposed the following changes that are relevant to conduct of operations associated with acceptance tests:

50 Change No.1: a modified basket design for the 24PTH DSC variants. As part of this change, the applicant increased the maximum allowed weight for the contents (fuel assembly and control components). As part of this change, the applicant modified the lead in both shield plugs of the 24PTH-S-LC DSC to allow for the use of precast lead as an option to the poured lead in the currently approved design. The modification includes differences in lead density and thickness and the introduction of radial gaps between the lead and the steel components of the shield plugs. For this change there were also associated changes to the dose rate limits in sections 3.2 and 3.3.2 of appendix A to the CoC. The staff also identified that other differences were proposed for application of dose rate limits to the different 24PTH DSC variants in the TC and the HSM.

12.1 Shielding Acceptance Tests UFSAR section P.9.1.5 includes the shielding acceptance, or integrity, tests for the NUHOMS storage system with the 24PTH DSCs. The current version of this UFSAR section, as provided with the amendment request submittal, does not recognize that the 24PTH-S-LC DSC shield plugs include lead and so does not provide specific shielding tests for the 24PTH-S-LC DSC shield plugs lead. The lead in the shield plugs provides significant axial shielding for this DSC.

Thus, appropriate acceptance tests are necessary to ensure the as-fabricated lead shielding, whether poured or precast, meets the minimum shielding performance requirements specified in the drawings and evaluated in the shielding analyses. While for steel components acceptance tests that are not specific to shielding (e.g., confirmation of compliance with requirements of material standards in the drawings, visual examinations, dimension measurements) are sufficient to demonstrate compliance with the drawings material and dimension specifications and thus ensure the shielding capability of the components, lead components need acceptance tests specific to their shielding capability.

Thus, the staff asked questions regarding acceptance tests for the 24PTH-S-LC shielding plugs lead components. In response to the staffs questions, the applicant revised UFSAR section P.9.1.5 to recognize the lead in the 24PTH-S-LC DSCs shield plugs and to provide acceptance tests with an associated acceptance criterion. The acceptance tests include two volumetric inspection method options: ultrasonic inspection, i.e., ultrasonic test, and gamma inspection, i.e., gamma scan. The staff finds these shielding acceptance test methods to be acceptable for the following reasons. First, these test methods have previously been identified by the staff as appropriate acceptance tests methods for lead shielding for spent fuel storage systems.

Second, the staff did not identify anything that would invalidate the use of these methods for the 24PTH-S-LC DSCs shield plugs lead.

The applicant proposed the following acceptance criterion: the effective lead thickness through any section conforms to the thickness specified in proprietary drawings NUH24PTH-1001-SAR (Rev 7A) and NUH24PTH-1002-SAR (Rev 3C). These drawings specify the minimum thicknesses of the lead components in the bottom and top axial DSC shield plugs for both the poured lead and the precast lead versions. For the precast lead, the drawings also specify a calculated density of 98.5% of the theoretical density. In evaluating the acceptability of the acceptance test criterion, the staff also looked at what is evaluated in the shielding analysis. The staff confirmed that the applicants shielding analysis used the minimum thicknesses specified in the design drawings. The staff also identified that the applicants analysis for the precast lead used a lead density that is 1.5% less than the density specified in the drawings. Because the acceptance test criterion is tied to the design drawing specifications and because the shielding analysis uses specifications or properties that slightly bound the drawing specifications, the staff finds the acceptance test criterion to also be acceptable.

51 Based on the above, the staff finds that the revised shielding acceptance tests in UFSAR section P.9.1.5 are acceptable and appropriate for ensuring the 24PTH-S-LC DSC meets the necessary specifications for adequate shielding performance.

12.2 Evaluation Findings F12.1 The staff finds that the acceptance tests for the NUHOMS 24PTH DSC systems are in compliance with 10 CFR Part 72 and that the applicable acceptance criteria have been satisfied. The evaluation of the acceptance tests and maintenance program provides reasonable assurance that the cask will allow safe storage of spent fuel throughout its licensed or certified term. This finding is reached on the basis of a review that considered the applicable regulations, appropriate regulatory guides, applicable codes and standards, and accepted practices.

13.0 QUALITY ASSURANCE EVALUATION The purpose of this review and evaluation is to determine whether TN has a quality assurance program that complies with the requirements of 10 CFR Part 20, Subpart G. The staff has previously reviewed and accepted the TN quality assurance program. There are no changes to the quality assurance program associated with the current application.

14.0 ACCIDENT ANALYSIS EVALUATION Accident analysis evaluations are performed in the SER technical sections 4.2.1.6, 4.2.2, 5.1.1, 6.1.3.1, 6.1.4.3, 6.2, 7.2.2.1, 8.0, 8.3, 8.4, and 10.1.

15.0 CONDITIONS FOR CASK USE - TECHNICAL SPECIFICATIONS The technical specifications and operating controls and limits review ensures that the operating controls and limits or the technical specifications, including their bases and justification, meet the requirements of 10 CFR Part 72. This evaluation is based on information provided in the application, as well as accepted practices, and the applicants commitments. For simplicity in defining the acceptance criteria and review procedures, the term technical specifications may be considered synonymous with operating controls and limits. The technical specifications define the conditions that are deemed necessary for safe dry storage system use. Specifically, they define operating limits and controls, monitoring instruments and control settings, surveillance requirements, design features, and administrative controls that ensure safe operation of the system.

15.1 Conditions for Use The conditions for use of the 24PTH DSC are clearly defined in the CoC and TS.

15.2 Standardized NUHOMS Certificate of Compliance Changes The Certificate of Compliance has been revised as follows:

Amendment No. 17 has been updated throughout to Amendment No. 18. Note that there was not an effective Amendment No. 12.

A missing space was added between 7160 and Riverwood Drive in the address on page 1.

Added a note below the table located in Section II.1.b Dry Shielded Canister (DSC) explaining that ASME B&PV Code,Section III, Division 1, Subsections NG and NF are

52 not applicable to the Type 3 basket assembly in 24PTH-S, 24PTH-L or the 24PTH-S-LC DSCs.

15.3 Changes to CoC Appendices Conforming changes to update the TS according to the changes associated with Amendment No. 18 are listed below. These changes are acceptable to staff.

Table 15-1 Conforming Changes to the CoC Appendices Cover pages The amendment level was changed to 18 in appendices A, B and C.

Tables of Content Added, removed or updated section titles as necessary as well as updated page numbers as necessary.

Appendix A, Section 2.0 Added basket type 3D for the 24PTH DSC Model to the table.

Appendix A, Section 2.4 Added new inspecting, testing and evaluation section for the high strength low allow steel used to fabricate the new basket design.

Appendix A, Section 3.1 Added Basket Types 1 and 2 to clarify which basket types are used with the 24PTH DSCs in high seismic storage modules.

Appendix A, Section 3.2 Consolidated the 24PTH-S and -L and the 24PTH-S-LC rows from the table Dose Rate Limits for the TC (except OS197L TC) into one entry, increased the value reported in the Axial Surface Dose Rate column for the 24PTH in the table Dose Rate Limits for the TC (except OS197L TC) and made the 24PTH-S-LC radial dose rate limit a footnote to the table Dose Rate Limits for the TC (except OS197L TC).

Appendix A, Section 3.3.2 Revised the HSM Door dose rate for the DSC Model 24PTH-S-LC and replaced 24PTH-S and -L with 24PTH in the DSC Model column of the table Dose Rate Limits for the Standardized HSM and HSM-H.

Appendix A, Section 4.2 Added criteria, e.g., cement specification type, water-cement ratio changes, etc., that would necessitate HSM concrete temperature testing.

Appendix A, Section 4.4 Deleted temperature testing requirements for newly loaded storage modules.

Appendix B, Section 2.1 Deleted statement that DSC models could be stored in accordance with appendix A, Inspections, Tests and Evaluations, section 4.4 Appendix B, Section 3.1.3 Removed the acronym DSC from all rows in the DSC Model column of the DSC transfer time limit completion table; added rows for the 24PTH-S-LC DSC with basket types 2A, 2B, 2C and 3D; added a row for the 24PTH-S and 24PTH-L DSCs with basket type 3D, added explanatory text to the existing Note about transfer time limits for 24PTH-S or 24PTH-L DSCs with Type 3 baskets and added a new Note below the Actions section of section 3.1.3.

Appendix B, Section 4.3.2 Moved all radiation protection text associated with the OS197L TC under the heading OS197L TC and moved NS text for all other TCs under the heading TCs with a Liquid Neutron Shield, Other Than the OS197L TC.

Appendix B, Section 4.3.6(b)

Added a table showing the maximum allowable air temperature rise through the HSM as a function of DSC and a clarifying note below the table.

53 Table 15-1 Conforming Changes to the CoC Appendices (contd)

Appendix B, Table 1-1e Removed a statement identifying the maximum allowable total weight of failed fuel cans from the Failed Fuel section and removed text implying end caps ensured retrievability from the Number and Location of Damaged Assemblies section.

Appendix B, Table 1-1l Removed a statement identifying the maximum allowable total weight of failed fuel cans from the Failed Fuel section, removed text implying end caps ensured retrievability from the Number and Location of Damaged Assemblies section, added the maximum allowable weight for a fuel assembly with control components for a Type 3D basket to the Maximum Assembly plus CC Weight section and added allowable decay heat limits associated with the Type 3D basket to the Decay Heat section.

Appendix B, Tables 1-1p thru 1-1q1 Added Type 3D basket to table heading.

Appendix B, Table 1-1r Added Type 3D basket and associated note to table.

Appendix B, Table 1-1t Removed a statement identifying the maximum allowable total weight of failed fuel cans from the Failed Fuel section and removed text implying end caps ensured retrievability from the Number and Location of Damaged Assemblies section Appendix B, Table 1-3i thru Table 1-3o, Table 1-7k and Table 1-7m Removed Maximum from the Table title.

Appendix B, Table 1-7m Deleted text at the bottom of page titled Notes for Tables 1-7k and 1-7m following Table 1-7m.

Appendix B, Figure 1-15 Revised Note 2 to reference Type 2 and Type 3 baskets versus Type 2A, 2B and 2C baskets Appendix B, Figure 1-15a Revised figure title to include Type 3 baskets Appendix C, Section C Added Type 1 and Type 2 to ASME Code Alternative Topic column of table and added Note 2 below table on page C-1.

Appendix C, Table C-8 Added Type 1 and Type 2 to table title.

Appendix C, Table C-12 Added text for ASME Code Section NG-4231.1.

15.4 Evaluation Findings Based on a review of the submitted information, the staff makes the following findings:

F15.1 The staff concludes that the conditions for use of the 24PTH DSCs with the Type 3D basket in conjunction with the Standardized NUHOMS System identify the necessary technical specifications to satisfy 10 CFR Part 72 and that the applicable acceptance criteria have been satisfied. The proposed CoC appendices provide reasonable assurance that the cask system will allow safe storage of spent fuel. This finding is based on the applicable regulations, appropriate regulatory guides, applicable codes and standards, and accepted practices.

54

16.0 CONCLUSION

The NRC staff has performed a comprehensive review of the application and has found that the Standardized NUHOMS System with the following changes meets the requirements of 10 CFR Part 72 (Note that change No. 5 and change No. 6 were withdrawn by the applicant):

Change No. 1:

Provide for a 24PTH improved basket design (Type 3) using staggered plates similar to EOS-37PTH to simplify construction, reduce weight and improve fabricability.

Change No. 2:

Deletion of Appendix A ITE requirement for initial HSM delta T measurement with a loaded DSC.

Change No. 3:

Clarifies Appendix B TS Section 4.3.2 language related to transfer casks with liquid neutron shields regarding the OS197L TC, which is significantly different than other TC models.

Change No. 4:

Updated Appendix C ASME Code Alternatives Table C-12 to add code alternative NG-4231.1 Change No. 7:

Clarify in Appendix B TS LCO 3.1.3 that there is no transfer time limit associated with the 24PTH-S-LC DSC, consistent with existing UFSAR analysis.

Change No. 8:

Editorially correct CoC name/address information by adding missing space between 7160 and Riverwood Drive.

Change No. 9:

Incorporate administrative controls during short duration ISFSI handling operations that are unanalyzed for tornado hazards in accordance with the guidance contained in NRC EGM 22-001, "Enforcement Discretion for Noncompliance of Tornado Hazards Protection Requirements at Independent Spent Fuel Storage Installations".

Additional Scope Change No. 1:

The applicant also requested, as an additional scope item, a change to the HSM concrete, to allow use of different cement, which is a blended Portland cement meeting the requirements of the ASTM C595 standard.

The areas of review addressed in NUREG-2215, Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities - Final Report 2020, are consistent with the applicants proposed changes. The certificate of compliance has been revised to include TNs requested changes. Based on the statements and representations contained in TNs application, as supplemented, the staff concludes that the changes described above to the approved contents of the Standardized NUHOMS System meet the requirements of 10 CFR Part 72.

Issued with Certificate of Compliance No. 1004, Amendment No. 18 on November 8, 2023.

v t e Letter Sequence Approval
CAC:001028, (Approved, Closed) EPID:L-2022-LLA-0079, Tn America, LLC, Application for Amendment 18 to Standardized NUHOMS Certificate of Compliance No. 1004 for Spent Fuel Storage Casks, Revision 0 (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance, Acceptance, Acceptance, Acceptance, Acceptance, Acceptance, Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting, Meeting Questions 25 August 2021 16 November 2018 4 February 2019 6 March 2019 14 March 2019 19 April 2019 23 April 2019 6 May 2019 7 January 2020 3 February 2020 6 March 2020 31 March 2020 31 March 2020 28 May 2020 4 August 2020 28 September 2020 6 October 2020 10 December 2020 23 March 2021 23 March 2021 15 June 2021 30 August 2021 28 October 2021 21 December 2021 20 January 2022 20 January 2022 31 March 2022 8 December 2022 4 January 2023 6 March 2023 6 March 2023 25 July 2023 25 July 2023 20 June 2023 4 October 2023 6 December 2023 22 November 2023 16 January 2024 20 March 2024 20 March 2024 20 May 2024 6 September 2024 9 October 2024 Answers 10 July 2020 11 September 2020 19 November 2021 30 November 2021 7 June 2022 29 July 2022 9 September 2022 3 October 2022 1 December 2022 21 December 2022 6 January 2023 27 January 2023 17 March 2023 8 May 2023 22 April 2024 4 November 2024 Results Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval, Approval Other: L-23-005, Independent Spent Fuel Storage Installation License Renewal Application, Revision 1 (Cac/Epid No. 001028/L-2022-RNW-0007), ML19150A559, ML19182A107, ML20002A442, ML20009D008, ML20010D564, ML20015A450, ML20015A451, ML20015A454, ML20028E845, ML20028E846, ML20049A082, ML20052E146, ML20052E154, ML20065N281, ML20071F156, ML20076D696, ML20099A180, ML20104C015, ML20104C018, ML20104C019, ML20105A135, ML20105A136, ML20105A137, ML20105A138, ML20105A139, ML20105A173, ML20111A240, ML20136A050, ML20136A051, ML20150A335, ML20150A421, ML20155K741, ML20155K743, ML20155K744, ML20161A088, ML20163A702, ML20163A703, ML20163A704, ML20181A211, ML20181A212, ML20192A328, ML20192A329, ML20212L768, ML20212L769, ML20261H420, ML20269A259, ML20279A577, ML20279A586, ML20289A493... further results
MONTHYEAR2019-07-01 [Table View]

ML23293A300

Text

SUMMARY

7.0 CRITICALITY EVALUATION

16.0 CONCLUSION

SUMMARY

1.7 Evaluation Findings

System Description

4.3 Evaluation Findings

5.4 Evaluation Findings

6.3 Evaluation Findings

7.0 CRITICALITY EVALUATION

7.3 Evaluation Findings

8.6 Findings

9.2 Evaluation Findings

16.0 CONCLUSION

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