Text

LLC Submittal of Topical Report Entitled NuScale US460 Fuel Storage Rack Design Topical Report, TR-145417-P, Revision 0
	ML23283A336
Person / Time
Site:	99902078, 05200050
Issue date:	10/09/2023
From:	Griffith T; NuScale
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23283A335	List: ML23283A336;
References
	LO-151775, TR-145417-P, Rev 0 TR-145417-NP, Rev 0
	Download: ML23283A336 (1)
	v • d • e

LO-151775 October , 2023 Docket No. 99902078 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Topical Report Entitled NuScale US460 Fuel Storage Rack Design Topical Report, TR-145417-P, Revision 0 NuScale Power, LLC (NuScale) hereby submits Revision 0 of topical report entitled NuScale US460 Fuel Storage Rack Design Topical Report, TR-145417-P. The purpose of this submittal is to request that the NRC review and approve the fuel storage rack design and analysis by October 2024. NuScale respectfully requests that the acceptance review be completed in 60 days from the date of transmittal. contains the proprietary version of the report entitled NuScale US460 Fuel Storage Rack Design Topical Report, TR-145417-P, Revision 0. NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavits (Enclosures 3 and 4) support this request. has also been determined to contain Export Controlled Information. This information must be protected from disclosure per the requirements of 10 CFR § 810. pertains to the NuScale proprietary information, denoted by double braces (i.e.,

((). Enclosure 4 pertains to the Framatome Inc. (formerly AREVA Inc.) proprietary information, denoted by brackets (i.e., [ ]). Enclosure 2 contains the nonproprietary version of the report. This letter makes no regulatory commitments and no revisions to any existing regulatory commitments. If you have any questions, please contact Kris Cummings at 240-833-3003 or at kcummings@nuscalepower.com. Sincerely, Thomas a Griffith Manager, Licensing NuScale Power, LLC Distribution: Matthew Mitchell, NRC Getachew Tesfaye, NRC Stacy Joseph, NRC NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

LO-151775 Page 2 of 2 10//23 Enclosure 1: NuScale US460 Fuel Storage Rack Design Topical Report, TR-145417-P, Revision 0, proprietary version Enclosure 2: NuScale US460 Fuel Storage Rack Design Topical Report, TR-145417-NP, Revision 0, nonproprietary version Enclosure 3: Affidavit of Carrie Fosaaen, AF-151777 Enclosure 4: Affidavit of Morris Byram, Framatome NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

LO-151775 : NuScale US460 Fuel Storage Rack Design Topical Report, TR-145417-P, Revision 0, proprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

LO-151775 : NuScale US460 Fuel Storage Rack Design Topical Report, TR-145417-NP, Revision 0, nonproprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Licensing Topical Report NuScale US460 Fuel Storage Rack Design Topical Report October 2023 Revision 0 Docket: 9992043 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com © Copyright 2023 by NuScale Power, LLC © Copyright 2023 by NuScale Power, LLC

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Licensing Topical Report REVISION HISTORY Revision Date Notes 0 October 2023 Initial Issuance © Copyright 2023 by NuScale Power, LLC

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Licensing Topical Report COPYRIGHT NOTICE This report has been prepared by NuScale Power, LLC and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC. The NRC is permitted to make the number of copies of the information contained in this report that is necessary for its internal use in connection with generic and plant-specific reviews and approvals, as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by NuScale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of these reports, the NRC is permitted to make the number of copies necessary for public viewing in appropriate docket files in public document rooms in Washington, DC, and elsewhere as may be required by NRC regulations. Copies made by the NRC must include this copyright notice and contain the proprietary marking if the original was identified as proprietary. © Copyright 2023 by NuScale Power, LLC

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Licensing Topical Report Department of Energy Acknowledgment and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008928. This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. © Copyright 2023 by NuScale Power, LLC

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table of Contents Abstract . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Executive Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.1 Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.2 Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1.3 Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.0 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 2.1 Regulatory Requirements, Codes, Standards, and Guidance . . . . . . . . . . . . . . . . . . . . . 7 2.2 Software . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.0 Design Description of Fuel Storage Racks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.1 Description of Spent Fuel Pool and Pertinent Fuel Parameters. . . . . . . . . . . . . . . . . . . 12 3.2 Description of Fuel Storage Racks. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 4.0 Structural Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4.1 Fuel Storage Rack Structural Model Development . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4.1.1 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4.1.2 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 4.1.3 Mesh Density Study and Benchmark Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . 25 4.2 Load Drop Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 4.2.1 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 4.2.2 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.2.3 Load Drop Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 4.3 [ ] . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 4.3.1 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 4.3.2 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 4.3.3 [ ] Results. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 4.4 Whole Pool Analysis of the Fuel Storage Rack Array in the Spent Fuel Pool . . . . . . . . 62 4.4.1 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 4.4.2 Whole Pool Analysis Modeling and Load Application . . . . . . . . . . . . . . . . . . . . 63 4.5 Fuel Storage Rack Stress Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 85 4.5.1 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 85 4.5.2 Service Level A/B Stress Results. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 4.5.3 Service Level D Stress Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 98 © Copyright 2023 by NuScale Power, LLC i

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table of Contents 4.5.4 Fuel Storage Rack Friction Loads Transferred to the Spent Fuel Pool Floor Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107 4.5.5 Rack-to-Rack Contact Forces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107 4.5.6 Maximum Reaction Forces on a Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . . 108 4.5.7 Rack Overturning Margin of Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 108 4.5.8 Rack Sliding Margin of Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 108 4.5.9 Whole Pool Analysis Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 111 5.0 Thermal-Hydraulic Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 114 5.1 Maximum Bulk Spent Fuel Pool Temperature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 114 5.1.1 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 114 5.1.2 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 114 5.1.3 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 114 5.2 Spent Fuel Pool Time-to-Boil Calculation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 115 5.2.1 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 116 5.2.2 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 117 5.3 Computational Fluid Dynamics Analysis to Establish Adequate Decay Heat Removal and Natural Circulation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 117 5.3.1 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 118 5.3.2 Modeling. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 119 5.3.3 Meshing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 122 5.3.4 Computational Fluid Dynamics Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 123 5.3.5 Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 125 6.0 Criticality Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 128 6.1 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 128 6.1.1 Depletion Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 128 6.1.2 Criticality Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 130 6.1.3 Calculation of the Maximum K . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 130 6.2 Assumptions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 131 6.3 Configuration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 132 6.3.1 Fuel Assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 132 6.3.2 Description of Fuel Storage System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 136 6.3.3 TRITON/ORIGEN-S Depletion Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 137 © Copyright 2023 by NuScale Power, LLC ii

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table of Contents 6.3.4 KENO-V.a Criticality Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 139 6.4 Initial Conditions, Boundary Conditions, and Limitations . . . . . . . . . . . . . . . . . . . . . . . 143 6.5 Analysis, Evaluation, and Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 144 6.5.1 Fuel Storage System Tolerance Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . 144 6.5.2 Fuel Assembly Physical Changes with Depletion . . . . . . . . . . . . . . . . . . . . . . 157 6.5.3 System Bias for Modeling Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 166 6.5.4 Fuel Storage Configurations and Assembly Eccentric Positioning Scenarios . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 174 6.5.5 Burnup and Enrichment Limits for Zone 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 188 6.5.6 Damaged Fuel Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 197 6.5.7 Assembly Dropped on Top of Rack . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 197 6.5.8 Burnup and Enrichment Selection for Accident Analysis . . . . . . . . . . . . . . . . . 198 6.5.9 Assembly Located Outside of the Fuel Storage Rack . . . . . . . . . . . . . . . . . . . 198 6.5.10 Single Misloaded Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 204 6.5.11 Multiple Misloaded Fuel Assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 204 6.5.12 Fuel Storage System Seismic Event . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 208 6.5.13 Temperature Variation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 211 6.5.14 Effect of Cooling Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 215 6.5.15 Neutron Spectrum . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 217 6.6 Summary of Criticality Evaluations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 218 6.6.1 Fuel Storage Rack Assembly Map . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 218 6.6.2 Definition of Assembly Enrichment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 218 6.6.3 Burnup Limit versus Assembly Enrichment . . . . . . . . . . . . . . . . . . . . . . . . . . . 218 6.6.4 Abnormal and Accident Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 219 6.6.5 Design Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 220 7.0 Materials Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 222 7.1 Material Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 222 7.2 Structural Material Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 223 7.3 Neutron Absorber Material Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 225 7.4 Material Evaluation Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 226 8.0 Manufacturing, Operation, and Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 227 © Copyright 2023 by NuScale Power, LLC iii

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table of Contents 8.1 Manufacturing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 227 8.2 Operations and Maintenance of the Fuel Storage Rack . . . . . . . . . . . . . . . . . . . . . . . 227 8.2.1 Neutron Absorber Monitoring Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . 227 8.2.2 Evaluating Neutron Absorber Test Results . . . . . . . . . . . . . . . . . . . . . . . . . . . 228 8.2.3 Neutron Absorber Coupons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 229 8.2.4 Neutron Absorber Coupon Tree . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 231 8.2.5 Spent Fuel Pool Water Chemistry Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . 231 8.2.6 Operations and Maintenance Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 231 9.0 Summary and Conclusions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 233 10.0 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 234 Appendix A Level D Member Stresses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-1 Appendix B Criticality Benchmark Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-1 B.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-1 B.2 Methodology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-2 B.2.1 Bias and Bias Uncertainty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-2 B.2.2 Normality . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-3 B.2.3 Expanded Shapiro-Wilk Test . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-3 B.2.4 ANSI N15.15 Normality Test . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-5 B.2.5 Non-Parametric Method . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-6 B.2.6 Trend Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-7 B.2.7 Determination of a Valid Trend. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-8 B.3 Selection of Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-9 B.3.1 Selection of Experiments for Bias and Uncertainty, Enrichment Trend, Plutonium Trend (with all cases) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-12 B.3.2 Selection of Experiments for Fuel Rod Pitch Trend . . . . . . . . . . . . . . . . . . . . .B-13 B.3.3 Selection of Experiments for Fuel Assembly Separation Trend . . . . . . . . . . . .B-14 B.3.4 Selection of Experiments for Soluble Boron Trend . . . . . . . . . . . . . . . . . . . . .B-15 B.3.5 Selection of Experiments for Boron Separator Plate Areal Density Trend . . . .B-15 B.3.6 Selection of Experiments for Moderator to Fuel Area Ratio Trend. . . . . . . . . .B-16 B.3.7 Selection of Experiments for Neutron Spectrum . . . . . . . . . . . . . . . . . . . . . . .B-17 B.3.8 Selection of Experiments for Plutonium Trend . . . . . . . . . . . . . . . . . . . . . . . . .B-18 B.3.9 Haut Taux de Combustion Experiments. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-18 © Copyright 2023 by NuScale Power, LLC iv

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table of Contents B.4 Results of Benchmark Calculations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-19 B.5 Trending Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-27 B.6 Test for Normality. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-55 B.7 Bias and Bias Uncertainty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-61 B.8 Benchmark Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-61 B.9 Implementation / Use . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-62 © Copyright 2023 by NuScale Power, LLC v

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 List of Tables Table 1-1 Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Table 4-1 [ ] Model Mesh Density Frequency Study Results . . . . . . . . . . . . . . 25 Table 4-2 [ ] Model Mesh Density Frequency Study Results . . . . . . . . . . . . . . . 25 Table 4-3 [ ] Model Mesh Density Stress Study Results . . . . . . . . . . . . . . . . . . 26 Table 4-4 Validation of Finite Element Model Mass and Center of Gravity . . . . . . . . . . . . 26 Table 4-5 Initial and Impact Velocities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 Table 4-6 Bounding Results Across Shallow and Deep Drop Scenarios . . . . . . . . . . . . . . 50 Table 4-7 Material Properties for a Fully Loaded Fuel Storage Rack. . . . . . . . . . . . . . . . . 56 Table 4-8 Whole Pool Analysis Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 Table 4-9 Level A & B Allowable Stress Criteria for Plate and Shell and Linear Type Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 Table 4-10 Level A & B Allowable Stress Criteria for Plate and Shell and Linear Type Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 87 Table 4-11 Level A & B Allowable Stress Criteria for Plate and Shell and Linear Type Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 89 Table 4-12 Maximum Design Level A Stresses [ ] . . . . . . . . . . 90 Table 4-13 [ ] . . . . . . . . . . . . . . . . . . . . . . . . 91 Table 4-14 Maximum Level A Stresses [ ] . . . . . . . . . . . . . . 92 Table 4-15 Weld Check, Service Level A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 93 Table 4-16 Maximum Leg Forces due to Deadweight Loading . . . . . . . . . . . . . . . . . . . . . . 95 Table 4-17 Maximum Design Level D Stress Ratios [

                               ] . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 100 Table 4-18         Maximum Level D Stress Ratios [                                                                    ] . . . . . . . . . 100 Table 4-19         Weld Check, Service Level D. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 102 Table 4-20         Maximum Support Leg Forces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 104 Table 4-21         Maximum Rack-to-Rack Impact Forces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107 Table 4-22         Maximum Horizontal and Vertical Forces on a Fuel Assembly . . . . . . . . . . . . 108 Table 4-23         Fuel Storage Rack Sliding, X-Direction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 109 Table 4-24         Fuel Storage Rack Sliding, Y-Direction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 110 Table 4-25         Fuel Storage Rack Design Stress/Interaction Ratio Summary. . . . . . . . . . . . . 112 Table 4-26         Fuel Storage Racks Critical Part and Weld Stress/Interaction Ratio Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 113 Table 5-1          Solver Under-Relaxation Factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 123 Table 5-2          Analysis Case and Peak Liquid Temperature . . . . . . . . . . . . . . . . . . . . . . . . . 126

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 List of Tables Table 6-1 Nuclides Credited in Depletion Calculation . . . . . . . . . . . . . . . . . . . . . . . . . . . 129 Table 6-2 Fuel Assembly Specification for Criticality Analysis . . . . . . . . . . . . . . . . . . . . . 133 Table 6-3 Material Properties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 135 Table 6-4 Fuel Storage Rack Design Parameters for Criticality Analysis. . . . . . . . . . . . . 136 Table 6-5 (( }}2(a),(c) . . . . . . . . 138 Table 6-6 Conservative Composite Axial Burnup Shapes . . . . . . . . . . . . . . . . . . . . . . . . 139 Table 6-7 Manufacturing Tolerance Uncertainty Factors versus Enrichment, 0 ppm Boron . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 145 Table 6-8 Manufacturing Tolerance Uncertainty Factors versus Enrichment, 1450 ppm Boron. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 146 Table 6-9 Combined Uncertainty Factors versus Enrichment, 0 ppm Boron . . . . . . . . . . 146 Table 6-10 Combined Uncertainty Factors versus Enrichment, 1450 ppm Boron . . . . . . . 147 Table 6-11 Combined Uncertainty Factor versus Enrichment and Burnup with No Soluble Boron. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 149 Table 6-12 Combined Uncertainty Factor versus Enrichment and Burnup with 1450 ppm Soluble Boron. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 150 Table 6-13 Standard Deviation of Combined Uncertainty Factor versus Enrichment and Burnup with No Soluble Boron . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 151 Table 6-14 Standard Deviation of Combined Uncertainty Factor versus Enrichment and Burnup with 1450 ppm Soluble Boron . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 153 Table 6-15 Final Burnup versus Enrichment Limit for Zone 1 . . . . . . . . . . . . . . . . . . . . . . 157 Table 6-16 Uncertainty Factors and Standard Deviation from Whole Pool Model . . . . . . . 157 Table 6-17 Spacer Grid and Pin Pitch Growth . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 158 Table 6-18 System Bias for Moderator Inside the Instrument and Guide Tubes . . . . . . . . 167 Table 6-19 Thermal Expansion Bias with No Soluble Boron . . . . . . . . . . . . . . . . . . . . . . . 167 Table 6-20 Standard Deviation of Thermal Expansion Bias with No Soluble Boron . . . . . 169 Table 6-21 Thermal Expansion Bias with 1450 ppm Soluble Boron . . . . . . . . . . . . . . . . . 170 Table 6-22 Standard Deviation of Thermal Expansion Bias with 1450 ppm Soluble Boron. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 171 Table 6-23 System Bias for Cladding Damage . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 173 Table 6-24 System Bias for Pin Pitch Growth . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 174 Table 6-25 Sensitivity Study of Eccentric Positioning for the Nominal Fuel Loading Map . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 188 Table 6-26 Dropped Fuel Assembly Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 203 Table 6-27 Single Misloaded Fuel Assembly Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . 204 © Copyright 2023 by NuScale Power, LLC vii

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 List of Tables Table 6-28 Multiple Misloaded Fuel Assembly Results . . . . . . . . . . . . . . . . . . . . . . . . . . . 208 Table 6-29 Summary of Seismic Event Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 209 Table 6-30 Seismically Induced Fuel Storage Rack Motion Results . . . . . . . . . . . . . . . . . 210 Table 6-31 [ ] . . . . . . . . . . . . . . . . . . . . . . . . . . . . 210 Table 6-32 Results for Simultaneous Fuel Storage Rack Shift [

                                    ] . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 210 Table 6-33         Spent Fuel Pool Reactivity versus Temperature Analysis . . . . . . . . . . . . . . . . 214 Table 6-34         Neutron Spectrum (Energy of Average Lethargy of Fission) . . . . . . . . . . . . . . 217 Table 6-35         Final Burnup versus Enrichment Limit for Zone 1 . . . . . . . . . . . . . . . . . . . . . . 219 Table 7-1          NuScale vs Typical Pressurized Water Reactor Spent Fuel Pool Water Chemistry Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 222 Table 8-1          Neutron Absorber Coupon Type and Count. . . . . . . . . . . . . . . . . . . . . . . . . . . 229 Table 8-2          Neutron Absorber Material in Spent Fuel Pool - Inspection Type and Schedule . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 230 Table B-1          Parameter Range for Critical Experiment Selection. . . . . . . . . . . . . . . . . . . . . .B-1 Table B-2          Critical Experiments Used for Bias Determination, Trending or Both. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-9 Table B-3          Benchmark Experiments Selected for Bias and Bias Uncertainty . . . . . . . . . .B-12 Table B-4          Benchmark Experiments Selected for Fuel Rod Pitch Trend . . . . . . . . . . . . . .B-13 Table B-5          Benchmark Experiments Selected for Fuel Assembly Separation Trend. . . . .B-14 Table B-6          Benchmark Experiments Selected for Soluble Boron Trend . . . . . . . . . . . . . .B-15 Table B-7          Benchmark Experiments Selected for Separator Plate Boron Areal Density Trend . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-16 Table B-8          Benchmark Experiments Selected for Moderator to Fuel Area Ratio Trend . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-16 Table B-9          Benchmark Experiments for Neutron Spectrum Trend . . . . . . . . . . . . . . . . . .B-17 Table B-10         Benchmark Experiments Selected for 240Pu Trend . . . . . . . . . . . . . . . . . . . .B-18 Table B-11         Critical Experiment Parameters and KENO V.a Results . . . . . . . . . . . . . . . . .B-20 Table B-12         Regression Analysis for Possible Bias Trending Variables . . . . . . . . . . . . . . .B-28 Table B-13         Data Used for Bias and Bias Uncertainty . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-56 Table B-14         Area of Applicability for Bias and Bias Uncertainty . . . . . . . . . . . . . . . . . . . . .B-62

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 List of Figures Figure 3-1 Fuel Storage Rack Array in the Spent Fuel Pool . . . . . . . . . . . . . . . . . . . . . . . . 14 Figure 3-2 General Layout of the Fuel Storage Rack . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 Figure 3-3 Fuel Storage Rack Without Storage Tubes, Lead-Ins, or Lead-In Perimeters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 Figure 3-4 Fuel Storage Rack Support Leg Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 Figure 4-1 Finite Element Model of the Fuel Storage Rack without Fuel Storage Tubes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 Figure 4-2 Simplified Base Model, Plan View of Fuel Storage Tube Modeling . . . . . . . . . . 22 Figure 4-3 Detailed Base Model, Plan View of Fuel Storage Tube Modeling . . . . . . . . . . . 22 Figure 4-4 Detailed Base Model, Plan View of Fuel Storage Tube Modeling . . . . . . . . . . . 24 Figure 4-5 Load Drop Scenario 1A - Shallow Drop, [ ] . . . . . . . . . . . . . . . . 35 Figure 4-6 Load Drop Scenario 1A - Local Mesh Sizing of Lead-Ins. . . . . . . . . . . . . . . . . . 36 Figure 4-7 Load Drop Scenario 1A - Local Mesh Sizing of Top Grid. . . . . . . . . . . . . . . . . . 36 Figure 4-8 Load Drop Scenario 1B - Shallow Drop, [ ] . . . . . . . . . . . . . . . . 39 Figure 4-9 Load Drop Scenario 1B - Local Mesh Sizing of Top Grid and Corner Post . . . . 40 Figure 4-10 Load Drop Scenario 1C - Shallow Drop, [ ]. . . . . . . . . . . . . . . . 42 Figure 4-11 Load Drop Scenario 1C - Local Mesh Sizing of Top Grid and Lead-Ins . . . . . . 43 Figure 4-12 Load Drop Scenario 2B - Deep Drop, [ ] . . . . . . . . . . . . . . . . . . 45 Figure 4-13 Load Drop Scenarios 2B and 2C - Plan View of Model . . . . . . . . . . . . . . . . . . . 46 Figure 4-14 Load Drop Scenarios 2B and 2C - Local Mesh Sizing of Baseplate . . . . . . . . . 47 Figure 4-15 Load Drop Scenario 2C - Deep Drop, [ ]. . . . . . . . . . . . . . . . . . 49 Figure 4-16 Plan View of the Fuel Storage Rack Array . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 Figure 4-17 Fuel Storage Rack and Spent Fuel Pool Symmetry Planes. . . . . . . . . . . . . . . . 54 Figure 4-18 [ ] Ground Motions and Associated Arias Intensities . . . . . . . . . . . . 58 Figure 4-19 Fuel Storage Rack Loading Considered in the Phase IV Analysis. . . . . . . . . . . 62 Figure 4-20 Detailed Fuel Storage Rack Finite Element Model. . . . . . . . . . . . . . . . . . . . . . . 64 Figure 4-21 Fuel Storage Rack and Spent Fuel Pool Finite Element Model . . . . . . . . . . . . . 65 Figure 4-22 Arrangement of Fuel Storage Racks in Spent Fuel Pool . . . . . . . . . . . . . . . . . . 66 Figure 4-23 Section of Fuel Storage Rack Model with Fuel Assemblies Loaded . . . . . . . . . 67 Figure 4-24 Fuel Storage Rack to Leg Plate Connection Model . . . . . . . . . . . . . . . . . . . . . . 68 Figure 4-25 Spent Fuel Pool Model Mesh, Fluid and Void (Left) and Concrete (Right) . . . . 70 Figure 4-26 Spent Fuel Pool Seismic Load Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 Figure 4-27 Contact Definitions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 74 Figure 4-28 Inner Fuel Tube Contact. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 75 © Copyright 2023 by NuScale Power, LLC ix

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 List of Figures Figure 4-29 Seismic Load Model and Concrete Tied Nodes . . . . . . . . . . . . . . . . . . . . . . . . . 75 Figure 4-30 Spent Fuel Pool Arbitrary-Lagrangian-Eulerian Interfaces. . . . . . . . . . . . . . . . . 77 Figure 4-31 Nodes for Seismic Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 79 Figure 4-32 [ ] Node 75634, Acceleration Time History, X-Direction . . . . . . . . . 80 Figure 4-33 [ ] Node 75634, Husid Diagram, X-Direction . . . . . . . . . . . . . . . . . . 81 Figure 4-34 [ ] Node 75634, Acceleration Time History, Y-Direction . . . . . . . . . 81 Figure 4-35 [ ] Node 75634, Husid Diagram, Y-Direction . . . . . . . . . . . . . . . . . . 82 Figure 4-36 [ ] Node 75634, Acceleration Time History, Z-Direction . . . . . . . . . 82 Figure 4-37 [ ] Node 75634, Husid Diagram, Z-Direction . . . . . . . . . . . . . . . . . . 83 Figure 4-38 Whole Pool Analysis, Smallest Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 84 Figure 4-39 Support Leg Layout . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 94 Figure 4-40 Rack 3, Top Inner Grid Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 103 Figure 4-41 Fuel Storage Racks, Leg 1, Sliding, X-Direction . . . . . . . . . . . . . . . . . . . . . . . 110 Figure 4-42 Fuel Storage Racks, Leg 1, Sliding, Y-Direction . . . . . . . . . . . . . . . . . . . . . . . 111 Figure 5-1 Spent Fuel Pool Heat Exchanger Concept. . . . . . . . . . . . . . . . . . . . . . . . . . . . 115 Figure 5-2 Top View of Computer Aided Design Model Geometry . . . . . . . . . . . . . . . . . . 120 Figure 5-3 Isometric View of Computer Aided Design Model Geometry . . . . . . . . . . . . . . 121 Figure 5-4 Bottom View of Computer Aided Design Model Mesh . . . . . . . . . . . . . . . . . . . 122 Figure 5-5 Top View of Spent Fuel Pool Showing Unique Active Fuel Assembly Zone Names . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 126 Figure 6-1 Fuel Pellet Chamfer and Shoulder Dimension (inches) with Tolerances . . . . . 134 Figure 6-2 Fuel Storage Rack Neutron Absorber Plate Layout for Revised Fuel Storage Racks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 137 Figure 6-3 KENO-V.a Model of a Single Fuel Storage Rack Location . . . . . . . . . . . . . . . 141 Figure 6-4 KENO-V.a Model of a Single Fuel Storage Rack Location Showing Corner Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 142 Figure 6-5 KENO-V.a Model of a Single Rack from the Whole Pool Model . . . . . . . . . . . 143 Figure 6-6 Nominal Fuel Loading Map (Map Option 5) . . . . . . . . . . . . . . . . . . . . . . . . . . . 155 Figure 6-7 Assembly Eccentric Position Map (Scenario 3) . . . . . . . . . . . . . . . . . . . . . . . . 156 Figure 6-8 Fuel Pellet Density versus Burnup . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 159 Figure 6-9 NEI 12-16 Appendix B Fuel Pellet Density versus Burnup. . . . . . . . . . . . . . . . 160 Figure 6-10 Fuel Pellet Diameter versus Burnup. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 161 Figure 6-11 NEI 12-16 Appendix B Fuel Pellet Diameter versus Burnup . . . . . . . . . . . . . . 162 Figure 6-12 Clad Outside Diameter versus Burnup. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 163 © Copyright 2023 by NuScale Power, LLC x

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 List of Figures Figure 6-13 NEI 12-16 Appendix B Clad Outside Diameter versus Burnup . . . . . . . . . . . . 164 Figure 6-14 Clad Thickness versus Burnup. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 165 Figure 6-15 NEI 12-16 Clad Thickness versus Burnup . . . . . . . . . . . . . . . . . . . . . . . . . . . . 166 Figure 6-16 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 6) . . . 175 Figure 6-17 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 7) . . . 176 Figure 6-18 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 8) . . . 177 Figure 6-19 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 9) . . . 178 Figure 6-20 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 10) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 179 Figure 6-21 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 11) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 180 Figure 6-22 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 12) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 181 Figure 6-23 Fuel Loading Map for Multiple Misloaded Fuel Assemblies (Map Option 13) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 182 Figure 6-24 Fuel Loading Map for Seismic Induced Rack Motion (Map Option 14) . . . . . . 183 Figure 6-25 Assembly Eccentric Positioning Map (Scenario 1). . . . . . . . . . . . . . . . . . . . . . 184 Figure 6-26 Assembly Eccentric Positioning Map (Scenario 2). . . . . . . . . . . . . . . . . . . . . . 185 Figure 6-27 Assembly Eccentric Positioning Map (Scenario 4). . . . . . . . . . . . . . . . . . . . . . 186 Figure 6-28 Assembly Eccentric Positioning Map (Scenario 5). . . . . . . . . . . . . . . . . . . . . . 187 Figure 6-29 k95/95 Versus Burnup for Fuel Enriched to 4.95 wt.% 235U . . . . . . . . . . . . . . . 189 Figure 6-30 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U . . . . . . . . . . . . 190 Figure 6-31 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U . . . . . . . . . . . . 191 Figure 6-32 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U . . . . . . . . . . . . 192 Figure 6-33 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U . . . . . . . . . . . . 193 Figure 6-34 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U . . . . . . . . . . . . 194 Figure 6-35 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U . . . . . . . . . . . . 195 Figure 6-36 Final Burnup versus Enrichment Limit for Zone 1 . . . . . . . . . . . . . . . . . . . . . . 197 Figure 6-37 Mislocated Fuel Assembly Location 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 199 Figure 6-38 Mislocated Fuel Assembly Location 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 200 Figure 6-39 Mislocated Fuel Assembly Location 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 201 Figure 6-40 Mislocated Fuel Assembly Location 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 202 Figure 6-41 Mislocated Fuel Assembly Vertical View . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 203 © Copyright 2023 by NuScale Power, LLC xi

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 List of Figures Figure 6-42 Keff versus Boron Concentration for Multiple Misloaded New Fuel Assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 206 Figure 6-43 Keff versus Boron Concentration for Multiple Misloaded Fuel Assemblies. . . . 207 Figure 6-44 Keff versus Spent Fuel Pool Temperature, 0 ppm Boron . . . . . . . . . . . . . . . . . 212 Figure 6-45 Keff versus Spent Fuel Pool Temperature, 1450 ppm Boron . . . . . . . . . . . . . . 213 Figure 6-46 Keff versus Decay Time After Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 216 Figure 6-47 Keff versus Decay Time After Shutdown, Details to 10 Days . . . . . . . . . . . . . . 217 Figure A-1 Corner Post Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-1 Figure A-2 Outer Brace Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-2 Figure A-3 Outer Fuel Tubes Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-2 Figure A-4 Inner Fuel Tube Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-3 Figure A-5 Bumper Web Shear Stress. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-3 Figure A-6 Bumper Flange Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-4 Figure A-7 Bottom Side Plate Shear Stress. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-4 Figure A-8 Baseplate Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-5 Figure A-9 Top Outer Grid Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-5 Figure A-10 Top Inner Grid Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-6 Figure A-11 Corner Post in Bottom and Top Grid Shear Stress . . . . . . . . . . . . . . . . . . . . . .A-6 Figure A-12 Bottom Outer Grid Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-7 Figure A-13 Bottom Inner Grid Shear Stress . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-7 Figure B-1 Regression Analysis of 235U Enrichment . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-29 Figure B-2 Histogram of Curve Fit Residuals for 235U Enrichment . . . . . . . . . . . . . . . . . .B-30 Figure B-3 Normality Plot of Curve Fit Residuals for 235U Enrichment . . . . . . . . . . . . . . .B-31 Figure B-4 Regression Analysis of Fuel Rod Pitch . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-32 Figure B-5 Histogram of Curve Fit Residuals for Fuel Rod Pitch. . . . . . . . . . . . . . . . . . . .B-33 Figure B-6 Normality Plot of Curve Fit for Residuals for Fuel Rod Pitch . . . . . . . . . . . . . .B-34 Figure B-7 Regression Analysis of Fuel Assembly Separation . . . . . . . . . . . . . . . . . . . . .B-35 Figure B-8 Histogram of Curve Fit Residuals for Fuel Assembly Separation . . . . . . . . . .B-36 Figure B-9 Normality Plot of Curve Fit Residuals for Fuel Assembly Separation . . . . . . .B-37 Figure B-10 Regression Analysis of Dissolved Boron Concentration . . . . . . . . . . . . . . . . .B-38 Figure B-11 Histogram of Curve Fit Residuals for Dissolved Boron Concentration. . . . . . .B-39 Figure B-12 Normality Plot of Curve Fit Residuals for Dissolved Boron Concentration. . . .B-40 Figure B-13 Regression Analysis of Separator Plate Areal 10B Density . . . . . . . . . . . . . . .B-41 © Copyright 2023 by NuScale Power, LLC xii

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 List of Figures Figure B-14 Histogram of Curve Fit Residuals for Separator Plate Area 10B Density . . . . .B-42 Figure B-15 Normality Plot of Curve Fit Residuals for Separator Plate Area 10B Density . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-43 Figure B-16 Regression Analysis of Moderator to Fuel Area Ratio . . . . . . . . . . . . . . . . . . .B-44 Figure B-17 Histogram of Curve Fit Residuals for Moderator to Fuel Area Ratio . . . . . . . .B-45 Figure B-18 Normality Plot of Curve Fit Residuals for Moderator to Fuel Area Ratio . . . . .B-46 Figure B-19 Regression Analysis of Neutron Spectrum. . . . . . . . . . . . . . . . . . . . . . . . . . . .B-47 Figure B-20 Histogram of Curve Fit Residuals for Neutron Spectrum . . . . . . . . . . . . . . . . .B-48 Figure B-21 Normality Plot of Curve Fit for Neutron Spectrum . . . . . . . . . . . . . . . . . . . . . .B-49 Figure B-22 Regression Analysis of 240Pu Enrichment (All Cases). . . . . . . . . . . . . . . . . . .B-50 Figure B-23 Histogram of Curve Fit Residuals for 240Pu Enrichment (All Experiments) . . .B-51 Figure B-24 Normality Plot of Curve Fit Residuals for 240Pu (All Experiments). . . . . . . . . .B-52 Figure B-25 Regression Analysis of 240Pu Enrichment (MOX Experiments). . . . . . . . . . . .B-53 Figure B-26 Histogram of Curve Fit Residuals of 240Pu Enrichment (Mixed Oxide Fuel Experiments) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-54 Figure B-27 Normality Plot of Curve Fit Residuals for 240Pu Enrichment (Mixed Oxide fuel experiments) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-55 © Copyright 2023 by NuScale Power, LLC xiii

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Abstract This licensing topical report contains a summary of the analytical inputs, interpretations, and methodologies used to design and analyze the Fuel Storage Racks (FSRs) to demonstrate compliance with applicable regulations in support the NuScale Power US460 Standard Plant Design. The FSRs are designed in accordance with the requirements of 10 CFR 50, Appendix B and ASME NQA-1. This report examines the FSRs in the areas of mechanical structure, thermal hydraulics, materials, and nuclear criticality. Structural analyses examine the mechanical integrity of the FSR design considering deadweight, impacts due to a dropped fuel assembly, and loading from a safe shutdown earthquake (SSE) in accordance with the Design Specific Review Standard, Section 3.8.4. Structural analyses additionally examine stresses due to lifting and the forces exerted in the event of a fuel assembly stuck in the rack. Thermal hydraulic analyses demonstrate that flow through the FSR is adequate for decay heat removal during anticipated operating occurrences and accident conditions and that natural circulation is sufficient to prevent nucleate boiling during anticipated operating conditions. Criticality analyses ensure that criticality control is maintained for normal and credible abnormal conditions in accordance with Standard Review Plan, Section 9.1.1. The analysis defines acceptable storage configurations for unirradiated and irradiated fuel that ensure criticality control is maintained. Evaluation of the structural and neutron absorbing materials assesses the chemical compatibility of the materials with the spent fuel pool environment. The analyses and evaluations demonstrate that the FSR design complies with applicable requirements and that the FSR design is acceptable for use in operating plants that use the US460 Standard Plant Design. This report also describes the quality programs applicable to manufacturing of the FSR and provides information concerning operation and monitoring of the FSRs. The operation and monitoring information provides the basis for development of FSR monitoring programs for licensees that adopt this report. © Copyright 2023 by NuScale Power, LLC 1

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Executive Summary This report contains a summary of the analytical inputs, methodologies, and results supporting the design and qualification of the FSR that is used to store new and spent fuel for the NuScale US460 Standard Plant Design. The information contained in this report demonstrates that the design of the FSR meets regulatory requirements in the areas of material selection, thermal-hydraulic performance, structural integrity, and nuclear criticality safety. The structural analysis of the FSR evaluates the structural integrity of the design considering deadweight, impact due to a dropped fuel assembly, and a SSE seismic event as described below: The fuel assembly drop is evaluated to the design criteria given in United States Nuclear Regulatory Commission (US NRC) Design-Specific Review Standard for NuScale SMR Design, Section 3.8.4, Other Seismic Category I Structures, Appendix D, Guidance on Spent Fuel Pool Racks, Item I.3: the functional capability of the FSR under impact loading is demonstrated. The FSR array in the Spent Fuel Pool (SFP) is evaluated to the design criteria given in US NRC Design-Specific Review Standard for NuScale SMR Design, Section 3.8.4, Appendix D, Item I.3: seismic analysis of the FSR array in the SFP considering SSE loading demonstrates that stresses in the racks meet criteria given in Subsection NF of Section III, Division 1, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code for Class 3 supports subjected to faulted loading. In addition, stresses due to lifting and the force exerted by the fuel handling machine in the event of a stuck fuel assembly are evaluated. The FSR in the Spent Fuel Pool (SFP) is shown to meet regulatory and applicable design code requirements related to structural performance of the FSR. The thermal-hydraulic analysis evaluates the ability to cool the spent fuel stored in the FSRs according to the design criteria given in US NRC Design-Specific Review Standard for NuScale SMR Design, Section 9.1.2, New and Spent Fuel Storage, Item III.4.I, and demonstrates that: Flow through the FSR is adequate for decay heat removal from the fuel assemblies during anticipated operating and accident conditions. There is adequate natural circulation of the coolant during anticipated operating conditions, including full core-offloads during refueling, to prevent nucleate boiling for the fuel assemblies. The FSR in the SFP is shown to meet regulatory requirements related to thermal-hydraulic performance of the FSR. © Copyright 2023 by NuScale Power, LLC 2

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The criticality safety analysis of the FSR and FSR array in the SFP evaluates the design to determine if criticality control is maintained per the criteria given in US NRC Standard Review Plan, Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and Handling, Item III.1, as described below: keff is demonstrated to be less than 1.0 for normal and credible abnormal conditions, considering relevant uncertainties and tolerances, when the FSRs are loaded with fuel of the maximum reactivity and the SFP is filled with full density unborated water. keff is demonstrated to be less than 0.95 for normal conditions, considering relevant uncertainties and tolerances, when the FSRs are loaded with fuel of the maximum reactivity and the SFP is filled with full-density water borated to a minimum concentration of 1450 ppm. The wetted parts of the neutron absorber material credited in the criticality calculations are shown to be compatible with, and chemically stable in, the SFP water. The design of the FSR and FSR array in the SFP precludes the placement of a fuel assembly between the FSRs. The criticality calculations also demonstrate that criticality limits are maintained if a fuel assembly is dropped between the FSRs and the SFP walls. The criticality analysis confirms acceptable locations for new and spent fuel and defines rules for governing the placement of the new and spent fuel in the FSRs. The FSR and FSR array in the SFP are shown to meet regulatory requirements related to the criticality safety. The materials analysis evaluates the structural and neutron absorbing materials used in the construction of the FSRs to determine if they are compatible with, and chemically stable in, the SFP water. In addition, a program for monitoring the effectiveness of the neutron absorber panels is described. The materials used in the fabrication of the FSR are shown to be acceptable and the neutron absorber monitoring program is shown to meet industry guidance. The evaluations, analyses, and neutron absorber monitoring program summarized within this document demonstrate that the FSR design and the configuration of the FSR array in the SFP meet the applicable regulatory requirements, design code requirements, and industry guidance governing the design of FSRs in SFPs. © Copyright 2023 by NuScale Power, LLC 3

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 1.0 Introduction 1.1 Purpose This topical report describes the design of the FSRs for use in the NuScale US460 Standard Plant Design. The FSRs are modular rack systems for storing, loading, and unloading fuel assemblies (FAs) in the SFP. Each FSR is a [

                                       ] to be safety stored in each of the FSRs in the SFP. The design incorporates fixed neutron absorbers. The racks are designed to store both irradiated and un-irradiated fuel in the SFP.

This report provides structural, thermal-hydraulic, material, and criticality analyses for the design of the FSRs. The analyses define the storage requirements for FA placement in the SFP. This report also describes manufacturing and monitoring requirements for the FSRs. Upon approval, this topical report will define the FSR design and provide the supporting safety analysis for the FSR for use with the US460 Standard Plant Design. 1.2 Scope This report defines the design of the FSR and describes the assumptions, codes, and methodology utilized to perform the structural, thermal-hydraulic, materials, and criticality analysis of the racks in the SFP. The elements of this report that are requested for approval are: The design of the FSRs. Acceptability for use of the FSRs in the spent fuel pool of plants that utilize the US460 Standard Plant Design. Suitability of the maintenance and operation test program description to support development of maintenance procedures for the FSRs. NuScale is not requesting generic approval for the methodologies used to perform the analyses presented in this report. 1.3 Abbreviations Table 1-1 Abbreviations Acronym Definition 3D 3-Dimensional AISI American Iron and Steel Institute ALE Arbitrary-Lagrangian-Eulerian ANS American Nuclear Society ANSI American National Standards Institute ASME American Society of Mechanical Engineers © Copyright 2023 by NuScale Power, LLC 4

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 1-1 Abbreviations (Continued) Acronym Definition ASTM American Society for the Testing of Materials B&PV Boiler and Pressure Vessel BC Boundary Condition CAD Computer Aided Design CFD Computational Fluid Dynamics CFR Code of Federal Regulations CG Center of Gravity CoP Coefficient of Performance CoR Coefficient of Restitution CRA Control Rod Assembly DSRS Design-Specific Review Standard EALF Energy of Average Lethargy of Fission EPRI Electric Power Research Institute EW East-West FA Fuel Assembly FCO Full Core Offload FEA Finite Element Analysis FEM Finite Element Model FSR Fuel Storage Rack GDC Generic Design Criteria GT IT Guide and Instrument Tube HTC Haut Taux de Combustion IC Ion Chromatography ID Inside Diameter IE Irradiation Embrittlement IR Interaction Ratio LMTD Log Mean Temperature Difference LWR Light Water Reactor MMC Metal Matrix Composite MOX Mixed Oxide (fuel) NEI Nuclear Energy Institute NPM Non-Parametric Margin NRC Nuclear Regulatory Commission NS North-South NSA Neutron Source Assembly NUREG Nuclear Regulatory Commission Technical Report Designation OBE Operating Basis Earthquake OD Outside Diameter OE Operating Experience PNNL Pacific Northwest National Laboratory PWR Pressurizer Water Reactor SCC Stress Corrosion Cracking SCW Site Cooling Water SDM Shutdown Margin SFP Spent Fuel Pool © Copyright 2023 by NuScale Power, LLC 5

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 1-1 Abbreviations (Continued) Acronym Definition SMR Small Modular Reactor SRSS Square Root of the Sum of the Squares SSE Safe-Shutdown Earthquake URF Under-Relaxation Factor WPA Whole Pool Analysis WRS Weld Residual Stress © Copyright 2023 by NuScale Power, LLC 6

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 2.0 Background Fuel storage racks provide a location for the storage of unirradiated and irradiated FAs that ensures there is no release of nuclear material during the long-term storage of the assemblies. Qualification of the FSR design addresses the following analyses Structural Analysis (Section 4.0)

             -    Load Drop Analysis
             -    [
                                                     ]
             -    WPA (deadweight, seismic, and stress analysis of the FSR)

Thermal-Hydraulic Analysis (Section 5.0)

             -    SFP Maximum Bulk Water Temperature Analysis
             -    SFP Time-to-Boil Analysis
             -    SFP Computational Fluid Dynamics (CFD) Analysis (to establish thermal conditions to be considered in the peak fuel cladding temperature analysis)
             -    Peak Fuel Cladding Temperature Analysis Criticality Safety Analysis (Section 6.0)
             -    Software Benchmark Analysis (Appendix B)
             -    Fuel Depletion Calculation
             -    Uncertainty Analysis
             -    FSR Reactivity Calculation Materials Analysis (Section 7.0)
             -    Assessment of Structural Materials
             -    Assessment of Neutron Absorber Material The information given for each analysis is sufficient to demonstrate compliance with applicable regulatory requirements, design codes and industry guidance. In addition to these analyses, Section 8.0 provides information on the monitoring program for the neutron absorber material.

2.1 Regulatory Requirements, Codes, Standards, and Guidance The design and qualification of the FSR meets the pertinent requirements of the General Design Criterion (GDC) given in Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix A, General Design Criteria for Nuclear Power Plants (Reference 10.3.b), through application of the guidance, recommendations, and requirements described in the regulatory guides and industry codes and standards that govern the materials, © Copyright 2023 by NuScale Power, LLC 7

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 thermal-hydraulic performance, structural integrity, and criticality safety of the FSR. It is noted that: [

Structural qualification of the FSR is in accordance with the 2017 Edition of Subsection NF and Appendix F to Section III, Division I, of the ASME B&PV Code (Reference 10.2.b.i). The quality requirements governing the design and qualification of the FSR are those given in Title 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants (Reference 10.3.c) and ASME NQA-1-2008, Quality Assurance Requirements for Nuclear Facility Applications, including NQA-1a-2009 (Reference 10.5). Applicable regulatory requirements, regulatory and industry guidance, and design code acceptance criteria applied in the design and qualification of the FSR are taken from the following documents:

1. 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, last updated August 28, 2007:

a. GDC 1, Quality Standards and Records

b. GDC 2, Design Bases for Protection Against Natural Phenomena

c. GDC 61, Fuel Storage and Handling and Radioactivity Control

d. GDC 62, Prevention of Criticality in Fuel Storage and Handling.

2. 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, last updated November 18, 2019.

3. 10 CFR 50.68, Criticality Accident Requirements, last updated November 16, 2006.

4. US NRC Regulatory Guide 1.29, Revision 6, Seismic Design Classification for Nuclear Power Plants, July 2021.

5. US NRC Regulatory Guide 1.61, Revision 1, Damping Values for Seismic Design of Nuclear Power Plants, March 2007.

6. US NRC Regulatory Guide 1.183, Revision 0, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.

7. US NRC Regulatory Guide 1.240, Revision 0, Fresh and Spent Fuel Pool Criticality Analyses, March 2021.

8. US NRC Design-Specific Review Standard Review for NuScale SMR Design:

a. Section 3.7.1, Revision 0, Seismic Design Parameters, June 2016.

b. Section 3.8.4 Revision 0, Other Seismic Category I Structures, June 2016.

c. Section 9.1.2, Revision 0, New and Spent Fuel Storage, June 2016.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

9. US NRC NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition

a. Section 9.1.1, Revision 3, Criticality Safety of Fresh and Spent Fuel Storage and Handling, March 2007.

10. US NRC NUREG/CR-6361, Criticality Benchmark Guide for Light-Water- Reactor Fuel in Transportation and Storage Packages, March 1997.

11. US NRC NUREG/CR-6665 (ORNL/TM-1999/303), Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel, February 2000.

12. US NRC NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, December 2000.

13. US NRC NUREG/CR-7109, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (keff) Predictions (NUREG/CR-7109, ORNL/TM-2011/514), April 2012.

14. ASME NQA-1-2008, Quality Assurance Requirements for Nuclear Facility Applications, including NQA-1a-2009 (2008 Edition with 2009 Addenda), August 2009.

15. ASME Boiler and Pressure Vessel Code, 2017 Edition, July 2017

a. Section II, Materials i) Part A, Ferrous Material Specifications ii) Part D, Properties (Customary)

b. Section III, Rules for Construction of Nuclear Facility Components, Division 1 i) Subsection NF, Supports ii) Nonmandatory Appendix F, Rules for Evaluation of Service Loadings with Level D Service Limits

c. Section V, Nondestructive Examination

d. Section IX, Welding and Brazing Qualifications.

16. American National Standards Institute (ANSI) / American Nuclear Society (ANS) 5.1-2014, Decay Heat Power in Light Water Reactors, reaffirmed February 2019.

17. ANSI/ANS-8.1-2014, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, reaffirmed June 2023.

18. ANSI/ANS-8.21-1995, Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors, reaffirmed April 2019.

19. ANSI/ANS-8.24-2017, Validation of Neutron Transport Methods for Nuclear Criticality Safety Calculations, reaffirmed January 2023.

20. ANSI N14.6-1993, Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4,500 kg) or More, January 1993.

21. ANSI N15.15-1974, NSI N15.15-1974, Assessment of the Assumption of Normality (Employing Individual Observed Values), October 1973 (Reaffirmed 1981).

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

22. Nuclear Energy Institute (NEI) 12-16, Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light Water Reactor Power Plants, September 2019.

23. American Institute of Steel Construction, Manual of Steel Construction, 15th Edition, 2017.

24. American Welding Society, AWS D1.6-1999, Structural Welding Code - Stainless Steel, 1999.

25. Electric Power Research Institute (EPRI), Pressurized Water Reactor Primary Water Chemistry Guidelines, Volume 1, Revision 7, 2014.

26. EPRI, Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175, Revision 1), 2017.

27. EPRI, Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Storage and Transportation Applications, Revision 1, 2022.

28. NEI 16-03-A, Revision 0, Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools, August 2016.

29. American Society for the Testing of Materials, ASTM G46-21, Standard Guide for Examination and Evaluation of Pitting Corrosion, August 2021.

2.2 Software Design and qualification of the FSR as described in this report utilizes analysis techniques that follow industry standard guidance and demonstrates that the applicable regulatory and design code requirements are met. Analyses supporting the design and qualification of the FSR are accomplished using industry standard commercially available software. The following software is used in the FSR design and qualification analysis. ANSYS Mechanical Version 19.2 (Reference 10.6) - ANSYS Mechanical is commercially available finite element modeling and analysis software for the solution of structural, thermal, acoustic, and non-linear physical problems. ANSYS Mechanical is used to develop FEMs of the FSR structure and to perform thermal-hydraulic analysis of the FSR. ANSYS LS-DYNA Version R10.1 (Reference 10.7) - ANSYS LS-DYNA is commercially available finite element modeling and analysis software for the explicit solution of dynamic loading such as drop tests, impact, and penetration. ANSYS LS-DYNA simulates the effects of high energy impacts as well as fluid-structure interaction. ANSYS LS-DYNA is used to perform the dynamic load drop analysis of the FSR and the seismic analysis of the FSR array in the SFP. ANSYS Fluent Version 2021/R2 - ANSYS Fluent is commercially available computation fluid dynamics software used to model fluid flow, heat and mass transfer, chemical reactions and more. ANSYS Fluent is used to perform thermal-hydraulic analysis of the SFP fluid system in order to predict temperatures and flow in the FSR environment. © Copyright 2023 by NuScale Power, LLC 10

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 SCALE Version 6.2.4 (Reference 10.8) - SCALE is a comprehensive modeling and simulation suite for nuclear safety analysis and design. It is used to perform analysis related to reactor physics, criticality safety, radiation shielding, spent fuel characterization, radioactive source term characterization, and sensitivity and uncertainty analysis for nuclear facilities and transportation and storage package designs. Criticality analysis of the FSR array is performed using KENO V.a, a three-dimensional Monte Carlo criticality computer code that is one of the primary criticality safety analysis tools in SCALE. The TRITON/ORIGEN-S module of SCALE is used to perform assembly depletion calculations to provided selected nuclide concentrations to KENO V.a for irradiated fuel. Commercially available software is qualified for use on safety-related tasks according to procedures governing the use and maintenance of this type of software. © Copyright 2023 by NuScale Power, LLC 11

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 3.0 Design Description of Fuel Storage Racks 3.1 Description of Spent Fuel Pool and Pertinent Fuel Parameters The design of the US460 plant includes a borated water SFP for the storage of fresh (unirradiated) and spent (irradiated) Fuel Assemblies (FAs). The FAs are stored in the SFP in a [

                                           ] in the FSR array are accessible by the fuel handling machine.

[ ] The design configuration of the FSR array in the SFP has the FSRs side-by-side [

                      ] Figure 3-1 shows the FSR array in the SFP.

The FSR design and the FSR layout in the SFP permit the storage of irradiated and fresh FAs with a maximum nominal enrichment of 4.95% wt.% 235U and an enrichment dependent minimum assembly average depletion as discussed in Section 6.0. The physical parameters of the stored fuel assemblies are as follows: Overall length (from the outside shoulders of the end fittings) is 94 inches. Interior cross section is 8.426 inches x 8.426 inches. Active fuel length is 78.74 inches. Dry weight without Control Rod Assembly (CRA) is [ ] Installation of the FSRs into the SFP is performed in any sequence deemed suitable during installation given that there is no fuel present in the SFP. [

3.2 Description of Fuel Storage Racks Each FSR is a [ ] assembly of square [ ] storage tubes (cells) with an inside dimension of [ ] inches on a center-to-center pitch of [ ] inches. The center-to-center pitch between two fuel storage cells in adjacent racks is approximately [ ] inches. The top edges of each storage cell are supplied with [ ] lead-ins that guide the FAs into the storage cell. The lead-ins are designed to allow identification markers to be embossed on them for tracking the locations of the stored FAs. The lead-ins are welded to each other along their top edges. © Copyright 2023 by NuScale Power, LLC 12

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 A single neutron absorber plate is sandwiched between each set of adjacent storage cells and a single neutron absorber plate is placed at the outside edge of each storage cell on the periphery of the FSR. In this manner the absorber material is placed on four sides of each storage cell, with the exception that the width of the absorber plates is reduced in the corners of the FSR in order to accommodate the corner posts. The neutron absorber plates extend beyond the active fuel length. The assembly of storage cells and neutron absorber plates is housed in a [

                ] external frame consisting of a baseplate, four corner posts, horizontal braces at three elevations, and top and bottom grid assemblies. The top and bottom grid assemblies provide lateral restraint to the storage tubes. The top grid assembly also provides vertical restraint to the storage tubes and neutron absorber plates. The baseplate supports the storage tubes and the four corner posts and is itself supported by eight adjustable support leg assemblies welded into holes on the underside of the baseplate, two in each quadrant. The baseplate is [                  ] thick in order to provide structural rigidity to the FSR and to prevent damage to the SFP liner in the event of an FA deep drop event. Figure 3-2 through Figure 3-4 provide the general layout of the FSR.

[

                                                                         ] Physical properties and allowable stresses are taken from Part D of Section II of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code (Reference 10.2.a.iii). The components are fabricated from a mixture of sheet steel, bar, and rolled steel stock of standard and readily available sizes. Standard material sizes are chosen to facilitate fabrication to the extent possible. The procurement and quality control procedures used in fabrication are in accordance with the requirements of Subsection NF of Section III, Division 1, of the ASME B&PV Code (Reference 10.2.b.i).

Structural welds used in the fabrication of the FSR are completed in accordance with Section IX of the ASME B&PV Code, Welding and Brazing Qualifications (Reference 10.2.d) using weld material with strength greater than that of the base material. © Copyright 2023 by NuScale Power, LLC 13

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 3-1 Fuel Storage Rack Array in the Spent Fuel Pool [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 3-2 General Layout of the Fuel Storage Rack [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 3-3 Fuel Storage Rack Without Storage Tubes, Lead-Ins, or Lead-In Perimeters [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 3-4 Fuel Storage Rack Support Leg Assembly [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.0 Structural Analysis The FSRs must demonstrate structural adequacy for their installed location in accordance with Appendix D, Guidance on Spent Fuel Pool Racks, of DSRS Section 3.8.4, Other Seismic Category I Structures (Reference 10.1.b). The following sections contain a summary of the analytical inputs, methodologies, acceptance criteria, and results for the following structural analyses. The load drop analysis evaluates the stresses in the FSR due to a dropped FA. [

Dynamic inputs used in the seismic analysis of the FSRs are described in Section 4.3. 4.1 Fuel Storage Rack Structural Model Development Two models of the FSR are created for use in the downstream load drop analysis and WPA using the commercially available ANSYS Mechanical (Reference 10.6) finite-element software package. [ ] model of the FSR is used in the WPA of the FSR array in the SFP. [ ] model of the FSR is the basis for the load drop analysis of the FSR and is modified for the load drop analysis as described in Section 4.2.1.2. These two models differ in the amount of detail they include, but both are shown to be representative of the key characteristics of the FSR important to the analyses for which they are used. 4.1.1 Methodology 4.1.1.1 Model Validation Approach A study is performed to validate that the mesh density used in the two base models is sufficient to adequately capture the response of the FSR. A [ ] model is created by dividing the elements in the [ ] model into four. The predominant frequencies of the two models are compared to validate the mesh density of the [ ] model. Predominant modes are those modes that have an effective mass to total mass ratio of at least [ ] The mesh density of the [ ] model is validated in a similar manner, through the creation of a [ ] model. The ability of the [ ] model to accurately predict stresses in the FSR is validated by comparing stresses obtained from the [ ] model loaded with fictitious static loads to stresses obtained from the [ ] model under the same loads. © Copyright 2023 by NuScale Power, LLC 18

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The [ ] model is created to reduce computer solution times for the load drop analysis. The predominant modal frequencies of the [

                          ] model are benchmarked against the predominant modal frequencies of the [                      ] model.

Both the [ ] models are also validated through a comparison of their weight and vertical Center of Gravity (CG), calculated by the finite-element software, to hand calculated values. Section 4.1.3 describes the mesh density study and the benchmarking analysis and results. 4.1.1.2 Modeling The modeling simplifications employed for both the [ ] and the [ ] models reduce the computer solution time without affecting the results of the subsequent analyses. These simplifications are described below. [

                                                                ] This results in [
                                      ] of the holes but improves the mesh shape of the bottom plate and connections with the vertical members.

At elevations between the [

[

                                                                                              . ] The neutron absorber plates serve no structural function and have negligible effect on the stiffness of the assembly given their relative flexibility.

[

                                                     ] serve no structural function and have negligible effect on the stiffness of the assembly. [
                                                          ] are modeled in the shallow load drop analysis described in Section 4.2.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [

The load drop analysis described in Section 4.2 models the [

                                                                          ] The WPA described in Section 4.4 considers this simplification when evaluating the local stresses.

[

In addition to the modifications described above, the [ ] model is modified as follows: [

The [ ] models are the same except in how the fuel storage tubes are represented. Figure 4-1 depicts the FSR model without the fuel storage tubes. Figure 4-2 and Figure 4-3 provide a plan view of how the fuel tubes are modeled for the [ ] model, respectively. [ ] are only used in the benchmark analysis described in Section 4.1.3. The WPA uses [

                                             ] definitions to model adjacent fuel tubes.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-1 Finite Element Model of the Fuel Storage Rack without Fuel Storage Tubes [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-2 Simplified Base Model, Plan View of Fuel Storage Tube Modeling [

Figure 4-3 Detailed Base Model, Plan View of Fuel Storage Tube Modeling [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.1.1.3 Material Properties Properties of the [ ] members included in the model are taken from Section II, Part D, of the ASME B&PV Code (Reference 10.2.a.iii), at the maximum normal condition operating temperature of the SFP, [ ] Modulus of Elasticity, E = 28 x 106 psi Poissons ratio, = 0.31 Density = 0.29 lb/in3 (except as modified by the simplifications described in Section 4.1.1.2). 4.1.1.4 Load Application The mesh density studies for the [ ] model require modal analyses to establish the predominant frequencies of the [ ] models that have a higher mesh density. The [ ] model is benchmarked to the [ ] model by comparing the predominant frequencies of the two models. Two fictitious static loads are applied to the [

                                   ] model and the resulting stresses from the two models are compared to demonstrate that the mesh density in the base model adequately predicts the stresses in the FSR. The first load applied, FA, is a lateral load with a magnitude of 100 kips and is applied to the [                                              ] at the halfway point of its length, as shown in Figure 4-4. The second load applied, FB, is a double lateral load of 50 kips each that are applied to the [
                                          ] at the halfway point of its length, as shown in Figure 4-4.

The mass of the [ ] are generated by applying the acceleration of gravity to the models. The mass of the models is validated through a comparison to hand calculated values. Similarly, the vertical CG of the two models is computed by the analysis software and is validated through comparison to the hand calculated value. © Copyright 2023 by NuScale Power, LLC 23

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-4 Detailed Base Model, Plan View of Fuel Storage Tube Modeling [

4.1.2 Acceptance Criteria The mesh density [ ] is considered acceptable if the frequencies of the predominant modes do not vary by more than [ ] versus the frequencies determined for the predominant modes of the [ ] model. The acceptance criteria used for the mesh study of the [ ] model is the same. The mesh density provided in the [ ] model is further validated if the stress intensities in the model due to the applied static loading do not vary by more than [ ] compared to those obtained from the [ ] model subjected to the same loading. The [ ] model is considered acceptable if the frequencies of the predominant modes do not vary by more than [ ] versus the frequencies determined for the predominant modes of the [ ] model. Both models are validated if the computer calculated mass and vertical CGs do not vary from hand calculated values by more than 5%. © Copyright 2023 by NuScale Power, LLC 24

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.1.3 Mesh Density Study and Benchmark Analysis The frequencies of each of the four models [

                                                   ] are determined in support of the mesh density study. The frequencies for the [                                                ] models are shown in Table 4-1 and Table 4-2, respectively. As indicated, the change in FSR frequency when comparing the lower mesh density base models and the higher density base models is well within the [             ] acceptance criteria described previously, indicating that the use of the base models in the follow-on analyses is appropriate.

Table 4-1 [ ] Model Mesh Density Frequency Study Results Frequency (Hz) Mode % Difference [ ] Bending in X direction [ ] Bending in Y direction [ ] Axial in Z direction [ ] Table 4-2 [ ] Model Mesh Density Frequency Study Results Frequency (Hz) Mode % Difference [ ] Bending in X direction [ ] Bending in Y direction [ ] Axial in Z direction [ ] Comparing the results of the [ ] model, a slight difference of -0.1% is observed horizontally. In the vertical direction, the [ ] model exhibits a higher natural frequency compared to the [ ] model. This discrepancy is acceptable as the fuel tubes in the [ ] model are based on an equivalent moment of inertia to that in the [ ] model, resulting in a stiffer vertical response. Given that the vertical response is well beyond the frequency range of interest, approximately 50 Hz, this difference is considered inconsequential. Two fictitious static loads are applied to the [

                                  ] model. Section 4.1.1.4 contains a description of the loads, and the resulting stress intensities at selected locations are compared in order to validate that the mesh density of the [                      ] model is sufficient to accurately predict stresses in the FSR. Stresses for the first load applied are taken at point A shown in Figure 4-4, which is at a point approximately one-third of a cell width. Stresses for the second load applied are taken at Point B shown in Figure 4-4, which is at a point approximately one-quarter of the bumper length. The stresses determined for these two load cases are shown in Table 4-3. As indicated, the change in stress is much smaller than the acceptance criterion of [               ] indicating that the use of the

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-3 [ ] Model Mesh Density Stress Study Results Stress Intensity (ksi) Load Application % Difference [ ] FA [ ] FB [ ] The mass and vertical CG of the [ ] model are determined by the analysis software. These values are compared to hand calculated values as another validation of the models. Table 4-4 shows the results of this comparison. As indicated, the mass and vertical CG of the models is very close to the hand calculated values, which is another indication that the models are appropriate for use in the downstream analyses. Table 4-4 Validation of Finite Element Model Mass and Center of Gravity Hand [ ][ ] Quantity Calculated Computer Computer Values % Difference % Difference Calculated Calculated Mass (lb-s2/in) [ ] 1 Vertical CG (in) [ ] 1 The vertical CG is measured from the center of the [ ] 4.2 Load Drop Analysis Table 1 and Section 4 of Appendix D to DSRS Section 3.8.4 (Reference 10.1.b), specify that the functionality of the FSR is demonstrated for the D + L + Fd load combination, where D is dead load acting on the FSR, L is the live load acting on the FSR and Fd is the force acting on the FSR due to the accidental drop of the heaviest load from the maximum possible height. 4.2.1 Methodology 4.2.1.1 Load Drop Scenarios The following load drop scenarios are postulated: Scenario 1A. Shallow drop, [

. ]

Scenario 1B. Shallow drop, [

. ]

Scenario 1C. Shallow drop, [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Scenario 2A. Deep drop, [

Scenario 2B. Deep drop, [

Scenario 2C. Deep drop, [

Scenario 3. Horizontal drop. [

4.2.1.2 Modeling The FSR models used in this analysis are based on the [ ] model described in Section 4.1. Descriptions of the modifications to the FSR model for each load drop scenario are provided in the sections below that discuss the individual load drop analyses. The resulting models are imported into LS-DYNA (Reference 10.7) for simulation of the various drop scenarios. [

                                       ] The [                            ] is therefore considered the heaviest load dropped on the FSRs. For the purposes of the load drop analysis, the weight of the [                                      ] which is slightly higher than the actual combined weight of these components and, therefore, conservative.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The dropped [ ] is modeled as a rigid body. This conservatively estimates damage to the FSR since the rigid body is not capable of absorbing kinetic energy while the actual [ ] would deform upon impact and divert some energy from the FSR. No stresses are calculated for the rigid body representing the [ ] [

                                ] Instead, a uniform gravity load is applied to the model elements and the dropped [              ] initial velocity is computed considering buoyancy forces.

This results in higher impact energy and is, therefore, conservative. 4.2.1.3 Material Properties Properties of the [ ] members included in the model are taken from Section II, Part D, of the ASME B&PV Code (Reference 10.2.a.iii), at the maximum operating temperature of the SFP, [ ] Modulus of Elasticity, E = 28 x 106 psi Poissons ratio, = 0.31 Density = 0.29 lb/in3 [

[ ] is the minimum tensile stress provided in Table NF-3324.5(a)-1 of Subsection NF to Section III of the AMSE B&PV Code (Reference 10.2.b.i). [

                               ] Consideration of engineering stress and strain is adequate for static and quasi-static loading conditions where plastic deformation is not expected.

[

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [

Section II, Part A, of the ASME B&PV Code (Reference 10.2.a.i) reports the minimum value of elongation at fracture for this material as [ ] It is assumed that the uniform strain of this material is [ ] of the elongation at fracture based on the character [

                         ] as shown in [
                                                                                         ] which yields:

[

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [

4.2.1.4 Load Application The velocity of the [ ] at impact with the FSR has two components: Initial velocity at a point just above the impact point on the FSR. This velocity is based on the maximum drop height of the [ ] and considers buoyancy of the [ ] in the SFP water in the load drop scenarios. [

Incremental velocity imparted by gravity as the [ ] accelerates from the point where the initial velocity is calculated to the FSR impact point. Table 4-5 provides the maximum drop height, initial velocity, and impact velocity calculated for each of the load drop scenarios. © Copyright 2023 by NuScale Power, LLC 31

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-5 Initial and Impact Velocities Maximum Drop Initial Velocity Impact Velocity Load Drop Scenarios Height(1) (in/s) (in/s) (in) 1A - Shallow Drop, [ ] 1B - Shallow Drop, [ ] 1C - Shallow Drop, [ ] 2B - Deep Drop, [ ] 2C - Deep Drop, [ ] 1 The Maximum Drop Height for the shallow load drop scenarios is the distance between the bottom of the [ ] at its maximum lift height above the FSR minus the height of the top of the FSR [ ] The Maximum Drop Height for the deep load drop scenarios is the distance between the bottom of the [ ] at its maximum lift height above the FSR minus the height of the top of the FSR [ ] The weight of the [ ] FSR is considered in the deep drop scenarios when calculating stresses in the support leg assembly: one-quarter of the [ ] FSR weight is considered along with the loads caused by impact. [

There are no live loads acting concurrently with the load drop scenarios. 4.2.2 Acceptance Criteria 4.2.2.1 General Acceptance Criteria Appendix D of Section 3.8.4 of the DSRS (Reference 10.1.b) [

                                     ] from Subsection NF to Section III, Division 1, of the ASME B&PV Code (Reference 10.2.b.i) for the D + L + Fd load combination. [
                                               ] is used to evaluate maximum stresses for those components that are required to maintain their configuration for the FSR to remain functional: [
                                                      ]

Stresses in the components of the FSR included in the analytical models are evaluated to these criteria to ensure they survive the [ ] impact loads. Stress limits for the full penetration welds are the same as the base metal. Bearing loads on the liner plate are transmitted to the SFP designer for evaluation. 4.2.2.2 Additional Shallow Drop Acceptance Criteria Per Table NF-3324.5(a)-1 and Table NF-3312.1(b)-1 of Subsection NF to Section III of the ASME B&PV Code (Reference 10.2.a.iii), the allowable shear stress in the [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [ ] Adjusting for the reduced section of the [

                                                                 ] If the calculated shear stress [
                         ] exceeds this criterion, results are considered acceptable if it is shown that the shear stress does not exceed the criterion [
                                                           ]

The [ ] are checked for buckling with the acceptance criteria being two-thirds of the ultimate engineering strength for the [ ] material from which these components are fabricated. The results of the analyses of the three shallow drop scenarios are also inspected to verify that the [ ] impact does not [

                                     ] affect the function of the neutron absorber plates and that the impact does not [
                                                              ] damage to the stored fuel assemblies. The gap between the bottom of the top grid and the top of the neutron absorber plate is [                 ] Therefore, the deflection of the [
                             ] at any time during impact must be less than [                           ]

the lateral deflection [ ] does not exceed the clearance between inside [

4.2.2.3 Additional Deep Drop Acceptance Criteria Stresses in the support leg assemblies, calculated with loads from the deep drop scenarios using classical stress formulas, are evaluated against the same stress criteria used when evaluating the FSR components included in the analytical models, as described in Section 4.2.2.1. The results of the analyses of the [ ] deep drop scenarios are inspected to verify that the [ ] impact does not result in [

                                                                         ] no contact is made with the SFP liner. The minimum gap between the [
                                  ] Therefore, maximum deflection in the [
                               ] In order to prevent penetration of the dropped [
                                 ] the stress and strain in the [                  ] must not exceed true ultimate stress or strain upon impact.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.2.3 Load Drop Analysis Results 4.2.3.1 Scenario 1A - Shallow Drop [ ] In this scenario, the [ ] drops vertically from its maximum height above the FSR and impacts the FSR [ ] For this analysis, the base model described in Section 4.1 is modified as follows: The mesh density is increased from the base model. Elements representing the [ ] that are not adjacent to the impact location are removed, leaving only those elements representing the [ ] that are nearest the impact location. Elements representing the [

                                      ] are removed.

Elements representing the [ ] included in the model are added using a sufficiently dense mesh (based on the mesh discretization study). The [ ] are set to their design value (the model described in Section 4.1 [

The mesh density [ ] in the vicinity of the impact location is increased. The nodes at the bottom [ ] are constrained in all six degrees of freedom. The rigid body representing the [ ] is added to the model in a vertical orientation and above the [ ] FSR where impact is simulated, with the bottom of the [ ] Removal of the elements that are not adjacent to the impact location reduces the computer solution time without significantly affecting stresses and deformations in those components most affected by the impact. As a result of these modifications, the model of the FSR for this scenario includes the [

The mesh density considered for the components nearest the impact location is validated through a mesh discretization study. The modeling approach taken provides a realistic contact relationship between the [ ] when impact occurs. Figure 4-5, Figure 4-6, and Figure 4-7 show the FSR model and dropped [ ] employed in this load drop scenario. © Copyright 2023 by NuScale Power, LLC 34

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-5 Load Drop Scenario 1A - Shallow Drop, [ ] [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-6 Load Drop Scenario 1A - Local Mesh Sizing of Lead-Ins [

Figure 4-7 Load Drop Scenario 1A - Local Mesh Sizing of Top Grid [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The worst-case impact location for Scenario 1A is an [

                                                          ] This provides the smallest impact area for a drop, giving the most conservative penetration and local stress results. It is assumed that the [              ] would distribute the impact load across the horizontal span of the [               ] if an FA is dropped closer to one corner. In this event, there would be a small reduction in local stress as well as in the total deformation of the [                                             ] Therefore, impact in the

[ ] is considered the most critical. For this load drop scenario, the bottom of the [ ] is initially assumed to be [ ] above the top of the [ ] with a calculated initial velocity of [ ]. Acceleration of gravity is applied to the entire model and the [ ] is accelerated into the FSR, resulting in a velocity at impact of [ ] that equates to a kinetic energy at impact of [ ] Two separate simulations are performed for this scenario: [

                                                                    . ] The true response of the FSR upon impact is bounded by these two simulations: the results from the case [
                                      ] provide the maximum stress and global deflection [
                         ] while the results for the case [                               ] provide the maximum penetration into the [                  ]

Analysis results show that the total energy in the system increases slightly over the duration of the event but remains stable, and that the hourglass energy remains a small fraction of the total energy, for both simulations (with and without element erosion). This energy profile indicates that the run is stable and that results are adequate for the purpose of this analysis. The analysis results also demonstrate that acceptance criteria are met for this load drop scenario, for both simulations. Table 4-6 contains a summary of the bounding results for load drop scenarios. 4.2.3.2 Scenario 1B - Shallow Drop [ ] In this scenario, the [ ] drops vertically from its maximum height above the FSR and impacts the FSR [ ] For this analysis, the base model described in Section 4.1 is modified as follows: The mesh density is increased from the base model. © Copyright 2023 by NuScale Power, LLC 37

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Elements representing the [ ] that are not adjacent to the impact location are removed, leaving only those elements [

                                              ] that are nearest the impact location.

Elements representing the lower grid, baseplate and support leg assemblies are removed. Elements representing the [ ] included in the model are added using a sufficiently dense mesh (based on the mesh discretization study). Elements representing [

                                             ] is added at the impact location using a sufficiently dense mesh. The [                 ] is removed from the FSR design.

The thickness of the elements representing the [ ] are set to their design value (the model described in Section 4.1 [

The mesh density [ ] in the vicinity of the impact location is increased. The nodes at the bottom [ ] are constrained in all six degrees of freedom. The rigid body representing the [ ] is added to the model in a vertical orientation and above [ ] where impact is simulated, with the bottom of the [ ] Removal of the elements that are not adjacent to the impact location reduces the computer solution time without significantly affecting stresses and deformations in those components most affected by the impact. As a result of these modifications, the model of the FSR for this scenario includes the [

The mesh density considered for the components nearest the impact location is validated through a mesh discretization study. The modeling approach provides a contact relationship between the [

                                                                                                      ] is removed from the design resulting in a larger impact velocity than modeled. The

[ ] is modeled [ ] above the top [ ] with an initial velocity of [ ] resulting in an impact velocity of [ ] upon impact [ ] Removal of the [ ] results in the [ ] and an impact velocity of [ ] This results in an increase of [ ] to the kinetic energy of the [ ] at impact. A [ ] increase in loading is deemed acceptable based on the analyzed condition Interaction Ratios (IRs) shown in Table 4-6. Figure 4-8 and Figure 4-9 contain the model as-analyzed. © Copyright 2023 by NuScale Power, LLC 38

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Analysis results show that the total energy in the system [ ] over the duration of the event but remains stable, and that the hourglass energy remains [ ] of the total energy. This energy profile indicates that the run is stable and that results are adequate for the purpose of this analysis. The analysis results also demonstrate that acceptance criteria are met for this load drop scenario. Table 4-6 contains a summary of the bounding results from the three shallow drop scenarios. Figure 4-8 Load Drop Scenario 1B - Shallow Drop, [ ] [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-9 Load Drop Scenario 1B - Local Mesh Sizing of Top Grid and Corner Post [

4.2.3.3 Scenario 1C - Shallow Drop [ ] In this scenario, the [ ] drops vertically from its maximum height above the FSR and impacts the FSR [ ] For this analysis, the base model described in Section 4.1 is modified as follows: The mesh density is increased from the base model. Elements representing [ ] that are not adjacent to the impact location are removed, leaving only those elements [

                                              ] that are nearest the impact location.

Elements representing the [

                                      ] are removed.

Elements representing the [ ] included in the model are added, using a dense mesh (based on the mesh discretization study). The thickness of the elements representing the [ ] are set to their design value (the model described in Section 4.1 [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The nodes at the bottom [ ] are constrained in all six degrees of freedom. The rigid body representing the [ ] is added to the model in a vertical orientation and above [ ] where impact is simulated, with the bottom of the [ ] Removal of the elements that are not adjacent to the impact location reduces the computer solution time without significantly affecting stresses and deformations in those components most affected by the impact. As a result of these modifications, the model of the FSR for this scenario includes the [

The mesh density considered for the components nearest the impact location is validated through a mesh discretization study. The modeling approach taken provides a realistic contact relationship [

                                                                       ] when impact occurs. Figure 4-10 and Figure 4-11 show the FSR model [                                      ] employed in this load drop scenario.

For this load drop scenario, the bottom of the [

                                                             ] with a calculated initial velocity of

[ ] Acceleration of gravity is applied to the entire model and the [ ] is accelerated into the FSR, resulting in a velocity at impact of [ ] that equates to a kinetic energy at impact of [ ] Analysis results show that the total energy in the system [ ] over the duration of the event but remains stable, and that the hourglass energy remains a [ ] of the total energy. This energy profile indicates that the run is stable and that results are adequate for the purpose of this analysis. The analysis results also demonstrate that acceptance criteria are met for this load drop scenario. Table 4-6 contains a summary of the bounding results from the three shallow drop scenarios. © Copyright 2023 by NuScale Power, LLC 41

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-10 Load Drop Scenario 1C - Shallow Drop, [ ] [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-11 Load Drop Scenario 1C - Local Mesh Sizing of Top Grid and Lead-Ins [

4.2.3.4 Scenario 2B - Deep Drop [ ] In this scenario, the [ ] drops vertically from its maximum height above the FSR [ ] and impacts the FSR [

                            ] above a [         ] support leg. For this analysis, the base model described in Section 4.1 is modified as follows:

The mesh density is increased from the base model. Elements above the [ ] are removed. Elements representing the [ ] are removed, and the [

Elements representing the [ ] are removed and the nodes defining the [ ] locations are constrained in all six degrees of freedom. The mesh density for those components that remain in the model is increased and the mesh density in the area of impact is further increased. The rigid body representing the [ ] is added to the model in a vertical orientation and above [ ] where impact is simulated, with the bottom of the [ ] © Copyright 2023 by NuScale Power, LLC 43

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Removal of the model elements [ ] reduces the computer solution time without significantly affecting the overall stiffness [

                               ] or the loads transferred to the support legs. As a result of these modifications, the model of the FSR for this scenario includes the [
                           ]

Constraint of [ ] locations increases the energy transferred, producing a conservative impact load for this deep drop scenario. Considering the minimal amount of [ ] vertical deformation that results from this analysis, this simplification does not significantly impact the conclusions drawn. The mesh increased density applied to this model is validated through a mesh discretization study. The modeling approach provides a realistic contact relationship [

                                              ] when impact occurs. Figure 4-12, Figure 4-13, and Figure 4-14 show the FSR model and dropped [                         ] employed in this load drop scenario.

For this load drop scenario, the bottom of the [ ] is initially assumed to be [ ] with a calculated initial velocity of [ ] Acceleration of gravity is applied to the entire model and the [ ] is accelerated into the FSR, resulting in a velocity at impact of [ ] Analysis results show that the total energy in the system [ ] over the duration of the event but remains stable, and that the hourglass energy [ ] of the total energy. This energy profile indicates that the run is stable and that results are adequate for the purpose of this analysis. The analysis results also demonstrate that acceptance criteria are met for this load drop scenario. Table 4-6 contains a summary of the bounding results from the two deep drop scenarios. © Copyright 2023 by NuScale Power, LLC 44

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-12 Load Drop Scenario 2B - Deep Drop, [ ] [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-13 Load Drop Scenarios 2B and 2C - Plan View of Model [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-14 Load Drop Scenarios 2B and 2C - Local Mesh Sizing of Baseplate [

4.2.3.5 Scenario 2C - Deep Drop [ ] In this scenario, the [ ] drops vertically from its maximum height above the FSR [ ] and impacts the FSR [ ] For this analysis, the base model described in Section 4.1 is modified as follows: The mesh density is increased from the base model. Elements above the [ ] are removed. © Copyright 2023 by NuScale Power, LLC 47

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Elements representing the [ ] are removed, and the [

Elements representing the [ ] are removed and the nodes defining the [ ] are constrained in all six degrees of freedom. The mesh density for those components that remain in the model is increased and the mesh density in the area of impact is further increased. The rigid body representing the [ ] is added to the model in a vertical orientation and above [ ] where impact is simulated, with the bottom of the [ ] Removal of the model elements [ ] reduces the computer solution time without significantly affecting the overall [

                                       ] loads transferred to the support legs. As a result of these modifications, the model of the FSR for this scenario includes the [
                           ]

Constraint of [ ] locations increases the energy transferred, producing a conservative impact load for this deep drop scenario. Considering the minimal amount of [ ] vertical deformation that results from this analysis, this simplification does not significantly impact the conclusions drawn. The mesh increased density applied to this model is validated through a mesh discretization study. The modeling approach provides a realistic contact relationship [

                                              ] when impact occurs. Figure 4-15 shows the FSR model and dropped [                ] employed in this load drop scenario. The [
                          ] is shown in Figure 4-13 and Figure 4-14.

For this load drop scenario, the bottom of the [ ] is initially assumed to be [ ] with a calculated initial velocity of [ ] Acceleration of gravity is applied to the entire model and the [ ] is accelerated into the FSR, resulting in a velocity at impact of [ ] that equates to a kinetic energy at impact of [ ] Analysis results show that the total energy in the system [ ] over the duration of the event but remains stable, and that the hourglass energy [ ] of the total energy. This energy profile indicates that the run is stable and that results are adequate for the purpose of this analysis. The analysis results also demonstrate that acceptance criteria are met for this load drop scenario. Table 4-6 contains a summary of the bounding results from the two deep drop scenarios. © Copyright 2023 by NuScale Power, LLC 48

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-15 Load Drop Scenario 2C - Deep Drop, [ ] [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.2.3.6 Load Drop Analysis Results Summary The load drop analysis results are presented in Table 4-6 below. Table 4-6 Bounding Results Across Shallow and Deep Drop Scenarios Result Interaction Drop Type Acceptance Criteria Limit (Scenario) Ratio [

[

[ Shallow Drop ] (Scenarios 1A/1B/1C) [

[

[ Deep Drop ] (Scenarios 2B/2C) [

[

1 Section 4.2.3.2 contains additional discussion on these results. 4.3 [ ] [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [

4.3.1 Methodology 4.3.1.1 Scenarios Considered The [ ] is performed in four phases: Phase I: Seismic input motions used in the US460 double building model are applied to two simplified models of the FSR array in the SFP. ((

                                      }}2(a),(c)

[

Phase II: The simplified model of the FSR array in the SFP and seismic case combination that provide the highest FSR acceleration in Phase I are utilized, [

Phase III: The simplified model of the FSR array in the SFP and seismic case combination that provide the highest FSR acceleration in Phase I are utilized, [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Phase IV: The simplified model of the FSR array in the SFP and seismic case combination that provide the highest FSR acceleration in Phase I are utilized, [

4.3.1.2 Modeling The FSR array in the SFP is represented by two simplified models: [

                                                                                 ] Plan views of the FSR array in the SFP showing these [                        ] are shown in Figure 4-16.

[

                                   ] In addition, both models have boundary conditions [
                                                                                                              ]

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-16 Plan View of the Fuel Storage Rack Array [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-17 Fuel Storage Rack and Spent Fuel Pool Symmetry Planes [

The SFP structure (walls and floor) is represented as a rigid body in both models. The opening on the east side wall of the SFP that allows pass-through of the FAs (the weir wall) [

                                                 ] 10 feet above the high water level. This places the elevation of the top of the SFP walls sufficiently higher than the high water elevation such that the fluid does not slosh out of the SFP. This is done to hold the water inventory within the SFP. In reality, water sloshing over the pool wall would be replaced through the weir wall such that the same volume of water would participate in hydrodynamic action due to seismic. ((
                           }}2(a),(c)

The SFP water is modeled in order to simulate fluid structure interaction between the water and the FSR as well as between the water and the SFP. Arbitrary Lagrangian Eulerian (ALE) elements from the ANSYS LS-DYNA (Reference 10.7) element library are used. © Copyright 2023 by NuScale Power, LLC 54

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [

The dry weight [ ] is assumed to be [ ] consistent with the weight used in the load drop and WPA analyses described in Section 4.2 and Section 4.4, respectively. [

                              ] However, since the analyses presented in this report are comparing results from an idealized rack, assuming that [
                                                    ] is acceptable.

The feet on the support legs are explicitly modeled and are connected to the underside of the baseplate [

                                                    ] The distance between the underside of the baseplate and the SFP floor liner is [                    ] the midpoint of the range allowed by the FSR design.

Contact between the FSRs and between the FSRs and the SFP (walls and floor) is modeled using automatic surface-to-surface contact with friction considered. The static coefficient of friction considered for FSR-to-FSR and FSR-to-SFP wall contact is [ ]. Static coefficients of friction for steel on steel vary from 0.74 to 0.78 under dry conditions and sliding coefficients of friction for steel on steel vary from 0.42 to 0.57 under dry conditions. A lower value of the coefficient of friction ( [ ] ) is considered for FSR-to-FSR and FSR-to-SFP wall contact because the FSR are submerged in water and the contact is thus lubricated. The coefficient of friction considered for FSR-to-SFP floor contact is varied from case-to-case as described in Section 4.3.1.1, with contact between all feet of a given FSR and the SFP floor liner having the same coefficient of friction. [

                                                                                                      ] between these two surfaces. This is a reasonable value for steel-on-steel contact that is based, in part, on the test results reported in National Aeronautics and Space Administration Technical Memorandum 106485, Influence of Temperature and Impact Velocity on the Coefficient of Restitution (Reference 10.13). [
                                                        ] Considering an FSR drop height of about

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [ ] the impact velocities for the racks is in the range of [ ] in air. The [ ] value is obtained by adjusting [

                                        ] by the ratio of velocities [            ]

4.3.1.3 Material Properties The material properties used to model the SFP structure are not relevant given that the structure is considered a rigid body for the purposes of this analysis, and it is therefore modeled using rigid elements. However, the properties assigned are representative of [ ] concrete. The properties of the SFP water are modeled with linear-polynomial equations of state using parameters of [

The FSR is modeled as linear-elastic material, with Youngs Modulus selected such that the first frequency of a single FSR model in-air is [ ] (which is approximately the first frequency determined from analysis of the detailed FSR base models as described in Section 4.1). As noted in Section 4.3.1.2, the density of the upper and lower halves of the 3D box representing the FSR [

                                                         ] Table 4-7 lists the material properties used for a fully loaded FSR. Youngs Modulus and densities for a partially loaded FSR are different.

Table 4-7 Material Properties for a Fully Loaded Fuel Storage Rack Property Value Mass Density, Top Half (lbf-s2/in/in3) [ ] Mass Density, Bottom Half (lbf-s2/in/in3) [ ] Youngs Modulus (psi) [ ] Poissons Ratio [ ] The FSR support legs are [

                                                                                           ] The feet themselves are essentially rigid [
                                                   ] Analysis results indicate that the feet are behaving

[ ] as expected. Therefore, material properties for the support legs and feet are not relevant. 4.3.1.4 Load Application Seismic motions in the form of acceleration time histories are applied to each of the SFP models in two orthogonal directions: for the NS Model, accelerations in the north-south (model y) direction and in the vertical (model z) direction are applied and for the EW Model ground accelerations in the east-west (model x) direction and in the vertical (model z) direction are applied. The applied © Copyright 2023 by NuScale Power, LLC 56

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 accelerations are those generated at [ ] The motion [ ] is used since the main driving force for the racks is the interaction between the support leg feet and the SFP floor. [

                                                                                                   ] The main effect of the seismic motions on the water is sloshing. However, the fluid velocities at the bottom of the pool, at the depth of the FSRs, are small given that the effect of sloshing reduces significantly with increasing depth.

The (( [ ] }}2(a),(c) ground acceleration time histories. The final analysis time is based on the Arias Intensity, a measure of the strength of the ground motion. The Arias Intensity, E, is calculated for the three orthogonal acceleration time histories for each of the seismic cases, as follows: tE 2 E = ³0 a ( t )dt Equation 4-11 where: t E = total earthquake duration a = earthquake acceleration The timepoint at which the Arias Intensity reaches [ ] is computed for each of the orthogonal directions and the largest of the three timepoints for a given seismic case is rounded up to the nearest 5 seconds (if it is not already a multiple of 5). This becomes the time at which the three orthogonal acceleration time histories for a given seismic case are truncated. The truncated acceleration time histories are used to determine the [ ] as described in Section 4.3.3. Figure 4-18 shows the three orthogonal ground acceleration time histories and their corresponding Arias Intensities for the [ ] earthquake, [

                                                                ] in Section 4.3.3. As indicated by the figure, the original orthogonal ground acceleration time histories are truncated at

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-18 [ ] Ground Motions and Associated Arias Intensities [

4.3.2 Acceptance Criteria There are no acceptance criteria associated with the [ ] analysis. 4.3.3 [ ] Results Analyses described in this section are executed using ANSYS LS-DYNA R10.1 (Reference 10.7). 4.3.3.1 Phase I Analysis Phase I analyses of the NS Model and EW Model are performed considering the (( }}2(a),(c) The initial FSR-to-FSR gap is set to [ ] (making the initial FSR-to-SFP wall gaps approximately [ ] or greater), the coefficient of friction between the FSR feet and the SFP floor liner is set at [ ] and the FSRs are [ ] The response of the FSRs is tracked in terms of the maximum FSR accelerations, the FSR-to-FSR and FSR-to-SFP wall contact forces, the final (post-earthquake) © Copyright 2023 by NuScale Power, LLC 58

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 distance between the FSRs, and the final distance between the SFP walls and the FSRs. The (( }}2(a),(c) are applied to the NS Model and the EW Model. The results are post-processed to determine the maximum accelerations, contact forces, final locations of the FSRs, and the minimum FSR-to-FSR and FSR-to-SFP wall gaps, in each model. These results indicate the amount of sliding experienced by the FSRs [

                                                                             ] Therefore, the

[

The Square Root of the Sum of the Squares (SRSS) of the accelerations in the three orthogonal directions is calculated for each FSR at each timepoint of the seismic loading cases. [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [

This approach is used to capture the response of the FSRs for a given seismic case in one number [ ] Examination of the (( [

                                                                   ] }}2(a),(c)

[

4.3.3.2 Phase II Analysis Phase II analyses consider four coefficients of friction [

                         ] between the FSR feet and the SFP floor liner. The [                 ] Model employed in Phase I (initial FSR-to-FSR gap of [                         ] and FSRs [
                                              ]) is revised to reflect the different FSR feet-to-SFP floor liner coefficients of friction. The bounding seismic case [
                                 ] is applied to the four Phase II NS Models and the results are post-processed to determine the maximum FSR accelerations and FSR-to-FSR contact forces. [
                                                                                                          ]

[

                                                                                                  ] Comparison of these four values shows that the results are not significantly different between the four cases. However, since the maximum acceleration of an FSR occurs when considering a coefficient of friction of [
                                                                    ]

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.3.3.3 Phase III Analysis Phase III analyses consider three initial FSR-to-FSR gaps [

                                                                                          ] The [        ] Model employed in Phase II (initial FSR-to-FSR gap of [                          ] FSRs [
                                              ] and an FSR feet-to-SFP floor liner coefficient of friction of

[ ]) is revised to reflect the revised initial FSR-to-FSR gaps of [

                                 ] The initial FSR-to-SFP wall gaps for the initial FSR-to-FSR gap of

[ ] are approximately [ ] or greater. The [ ] seismic case [ ] is applied to the two new Phase III [ ] Models, and the results are post-processed to generate maximum FSR accelerations and FSR-to-FSR contact forces. The post-processed results from these three cases are compared to determine the maximum FSR accelerations and FSR-to-FSR contact forces. [

[

                                                                                              ] Comparison of these three values indicates that the results are not significantly different between the three cases. [
                                                                                  ]

4.3.3.4 Phase IV Analysis Phase IV analyses consider a case with [

                                                                                               ] as shown in Figure 4-19. The [           ] Model employed in Phase III (initial FSR-to-FSR gap of

[ ], FSRs [ ] and FSR feet-to-SFP floor liner coefficient of friction of [ ]) is revised to reflect [

                             ] The [                ] seismic case [                                               ]

is applied to the new Phase IV [ ] Model and results are post-processed to generate maximum FSR accelerations and FSR-to-FSR contact forces. The post-processed results from this case and the equivalent Phase III case with fully loaded FSRs are compared to determine the maximum FSR accelerations and FSR-to-FSR contact forces. [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-19 Fuel Storage Rack Loading Considered in the Phase IV Analysis [

[

4.4 Whole Pool Analysis of the Fuel Storage Rack Array in the Spent Fuel Pool 4.4.1 Methodology Based on the findings of the [ ] analysis (Section 4.3), the WPA analysis is performed using the [ ] earthquake with [ ] FSRs placed [ ] and a friction coefficient of [ ] between the FSR feet and the SFP floor liner. The seismic interaction of the FSRs within the SFP is simulated using a FEM to determine the seismic forces acting on the FSRs during a SSE event. This requires consideration of the hydrodynamic coupling between the FSRs and fluid within the SFP. The whole pool seismic analysis is performed using the LS-DYNA Finite Element Analysis (FEA) code. The WPA uses the ALE capability in LS-DYNA. The FSRs are modeled with a Lagrangian mesh where the coordinates of the nodes move with the material of the FSRs. The fluid is modeled with an ALE mesh where fluid material advects or flows through the mesh. The ALE mesh is defined to follow the motions of the SFP walls and floor with the ALE reference mesh. Void spaces are modeled where fluid elements can flow during the analysis. A void space is continuously connected with the fluid only at the head space above the top water level. The void space is initially absent of fluid material. The fluid-structure interaction is defined between the Lagrangian structures (the FSRs) and the ALE structures (the © Copyright 2023 by NuScale Power, LLC 62

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 fluid) via the constrained Lagrange in solid boundary definition. The FSRs and the FAs act as slave segments to ALE masters for the contact definition. The analysis is run in accordance with Appendix D to Section 3.8.4 of the DSRS (Reference 10.1.b) to determine FSR design stresses due to seismic loading, FA impact/drop, and rack-to-rack interaction. The model used in the WPA includes detailed representations of the [ ] FSRs, the SFP fluid, the void space (air) above the fluid, and the concrete SFP walls and floor. This model is referred to herein as the WPA Model. The WPA Model is analyzed for one set of time history input motions, considering a coefficient of friction [ ] between the FSRs (leg plates) and SFP floor liner. The steps performed in the WPA are:

1. The detailed FSR model is developed using ANSYS (Section 4.1) for use in the WPA Model.

2. Multiple FSRs, concrete SFP, fluid, and void are modeled. The SFP is modeled as a 3D rectangular tank with [ ] submerged detailed FSRs in ANSYS.

3. The ANSYS finite element mesh of the WPA Model is then converted to LS-DYNA input (*.inc files).

4. The WPA is performed using the program LS-DYNA R10.1 (Reference 10.7).

Additional material, contact, and part definitions are added to the LS-DYNA model. Appropriate friction parameters are added between the FSR leg plates and SFP floor liner.

5. Gravity, buoyancy, and seismic accelerations are applied to the LS-DYNA model.

Multi-point seismic excitations in three orthogonal directions are applied simultaneously.

6. The results of the WPA are post-processed to determine the following:

                  -   Maximum FSR stresses
                  -   Maximum floor reaction under one leg
                  -   Maximum reaction on an FA due to interaction with the FSR Utilizing the computed seismic loads, which include dead loads, FSR design is performed. The FSR design also considers results from the load drop analysis described in Section 4.2 and the thermal analysis described in Section 5.0.

4.4.2 Whole Pool Analysis Modeling and Load Application 4.4.2.1 Fuel Storage Rack and Fuel Assembly Model Development An ANSYS model of the FSR is developed as described in Section 4.1 and is shown in Figure 4-20. This model is repeated [ ] FSR assemblies in the WPA Model as shown in Figure 4-21. The FSRs are modeled © Copyright 2023 by NuScale Power, LLC 63

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [ ] between the bumper plates. The FSRs are labeled [ ] as shown in Figure 4-22. Figure 4-23 shows the model with the FAs loaded [ ] The FAs are modeled with solid elements. The FA has lateral natural frequencies of [ ] for beginning-of-life and end-of-life conditions, respectively. [

                                                                                          ] Physical modeling of the FAs allows for the evaluation of the FA potential interaction with adjacent tube walls under seismic input motions.

Figure 4-20 Detailed Fuel Storage Rack Finite Element Model [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-21 Fuel Storage Rack and Spent Fuel Pool Finite Element Model [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-22 Arrangement of Fuel Storage Racks in Spent Fuel Pool [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-23 Section of Fuel Storage Rack Model with Fuel Assemblies Loaded [

4.4.2.2 Fuel Storage Rack Mesh Conversion from ANSYS to LS-DYNA The node and element connectivity of the FSR model is converted from ANSYS into LS-DYNA format. The FSR shell elements are modeled using seven integration points and the Belytschko-Wong-Chiang formulation (ELFORM=10), which is the same as the default but with better mitigation of warped area configurations. FA solid elements are fully integrated solid elements (ELFORM = 2) with hourglass control IHQ = 5 (Flanagan-Belytschko stiffness form with exact volume integration). [ ] material properties [ ] are assigned to the FSR shell elements utilizing the [ ] material model [ ] The FSR shell materials use a Poissons ratio = 0.31, an engineering yield stress [ ] and a Tangent Modulus © Copyright 2023 by NuScale Power, LLC 67

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [ ] For the FAs an [ ] material model is used [ ] The leg plates are connected to the [

                                             ] using the LS-DYNA command

[ ] Figure 4-24. Therefore, the FSR legs are modeled as [

Figure 4-24 Fuel Storage Rack to Leg Plate Connection Model [

4.4.2.3 Spent Fuel Pool Model Development The SFP model is developed where the bottom slab and top-of-the-wall elevations are [ ] respectively. The inside SFP lateral dimensions are [ ] The SFP water top elevation is [ ] Therefore, the total height of the water is [ ] To allow for possible sloshing, the SFP walls are modeled [ ] above the top of the water. An ANSYS mesh of the SFP with fluid, void, and concrete structure is generated. To perform the fluid-structure interaction analysis, water and voids are meshed with ALE domain solid elements. The concrete basin is meshed using Lagrangian shell elements with node-to-node connectivity between fluid and © Copyright 2023 by NuScale Power, LLC 68

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 basin. The ALE domain moves with the Lagrangian basin concrete structure and is thus subjected to the seismic motions as well. The detailed FSR model targets an element size of [ ] To achieve an optimal fluid-structure interaction the fluid elements in the bottom part of the SFP are also modeled at a size of [ ] Above the FSR module the fine mesh is transitioned into a coarser mesh with an element size of approximately [ ] An air void is modeled above the top of the water level to allow for potential sloshing. The SFP concrete structure boundary consists of the floor slab and four side walls. The North side wall is at grid line RX-C, the East wall at grid line RX-2.4, the South wall at grid line RX-D, and the West wall at grid line RX-2 (Figure 4-22). The concrete structure is modeled to hold the fluid as shown in Figure 4-25. Concrete, fluid, and voids are modeled as a continuous mesh. The concrete structure is modeled using shell elements. Since the fluid and void outside nodes correspond to the inside faces of the SFP the concrete shell elements need to be offset. The concrete structure shell elements are oriented such that the element normals point away from the fluid resulting in positive offsets. [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-25 Spent Fuel Pool Model Mesh, Fluid and Void (Left) and Concrete (Right) [

4.4.2.4 Spent Fuel Pool Mesh Conversion from ANSYS to LS-DYNA The SFP (walls, fluid, and void) model is converted from ANSYS into LS-DYNA format. Concrete shell elements are modeled using two integration points (no stress output needed) and the fully integrated shell element formulation (ELFORM=16). SFP fluid and void solid elements are one integration point ALE solid elements (ELFORM = 12) with hourglass control IHQ = 1 (viscous form). The fluid and void materials use the mat-null material model [

                                                        ] The properties of water at the operating temperature [            ] and atmospheric pressure are Water weight density: [                                       ]

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Water mass density: [ ] Water bulk modulus: [ ] For the SFP concrete structure an [ ] material model is used [ ] A normal weight concrete with a compressive strength of [ ] is used. The properties of the concrete are Concrete weight density: [ ] Concrete mass density: [ ] Concrete Modulus of Elasticity: [ ] The ALE reference coordinate system is issued with reference to nodes of the SFP seismic load model (Section 4.4.2.3) such that the ALE parts follow the input motions (defined with *ALE_REFERENCE_SYSTEM_NODE). 4.4.2.5 Spent Fuel Pool Seismic Load Model Development Seismic motions in the form of acceleration time histories in three spatial directions are provided for 832 nodes. A [ ] concrete layer is developed from the nodal coordinates (Figure 4-26) of these nodes. Concrete shell elements are modeled using two integration points (no stress output needed) and the fully integrated shell element formulation (ELFORM=16). © Copyright 2023 by NuScale Power, LLC 71

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-26 Spent Fuel Pool Seismic Load Model [

4.4.2.6 Whole Pool Analysis Model The WPA Model is generated by combining the SFP seismic load model from Section 4.4.2.5, the SFP model from Section 4.4.2.3, and the FSR model from Section 4.4.2.1 into one FEM. Constant damping proportional to element deformation over a specified frequency range is included in the WPA model. The damping is applied to the model utilizing the *DAMPING_FREQUENCY_RANGE_DEFORM command. Per NRC Regulatory Guide 1.61 (Reference 10.14, steel members of the FSR (except leg plates) have [ ] The first natural frequency of the FSR [ ], which defines the first frequency for the frequency range damping definition. The upper bound frequency for this © Copyright 2023 by NuScale Power, LLC 72

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 analysis is set to [ ] defining the end of the frequency range. The FAs have natural frequencies starting at [

                                      ] The only concrete wall that is not constrained by seismic motion is the East wall where a damping [               ] is used. Table 4-8 provides a summary of WPA model damping definitions.

Table 4-8 Whole Pool Analysis Damping Lowest Highest Description Damping Frequency [Hz] Frequency [Hz] FSRs w/o leg plates [ ] FAs [ ] East wall [ ] 4.4.2.7 Contact Definitions [

The coefficient of friction for the contact definitions is [ ] (for steel on steel). Contact between the leg plates and the floor are automatic surface-to-surface contacts and are used for the contact between the leg plates and the seismic load model. Each leg plate is defined to have contact with the entire SFP floor (the contact definition is such that each leg plate can contact the SFP floor at any point). The leg plates [ ] is specified in the contact surface to avoid undesirable oscillation. The coefficient of friction between leg plate and floor (with steel liner) is [ ] The seismic load model that is used for the load application is tied to the concrete structure using the tied nodes to surface contact card (Figure 4-29). © Copyright 2023 by NuScale Power, LLC 73

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-27 Contact Definitions [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-28 Inner Fuel Tube Contact [

Figure 4-29 Seismic Load Model and Concrete Tied Nodes [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.4.2.8 Interface Dimensions There are three SFP ALE interfaces: the interface between fluid and void, the interface between fluid and concrete structure, and the interface between void and concrete structure. These interfaces are shown in Figure 4-30 and are realized through continuous modeling (nodes are shared at the interface). The interface between the FSRs and FAs and the fluid is modeled using the LS-DYNA command *CONSTRAINED_LAGRANGE_IN_SOLID. The FSR shell elements, except those representing the bumpers, as well as the outside faces of the FA solid elements are included in the CONSTRAINED_LAGRANGE_IN_SOLID definition. A penalty base coupling method is utilized in the element normal direction (compression and tension) only (no friction). The FSR Lagrangian shell elements have approximately the same size as the ALE elements. With that, the number of additional coupling points on the Lagrangian elements is set to two points per element width (NQUAD = 2) resulting in a total of four coupling points. © Copyright 2023 by NuScale Power, LLC 76

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-30 Spent Fuel Pool Arbitrary-Lagrangian-Eulerian Interfaces [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.4.2.9 Load Application There are three external load types acting on the system: gravity, buoyancy, and seismic accelerations that are applied over a defined time period. Gravity is applied to the model as a body force through a base acceleration. The acceleration for gravity in the model is 386.4 in/s2 [

                                ] The buoyant force of the FSR is calculated based on the assumption that the entire FSR is made from the same [
                         ] material. The exceptions are the neutron absorber plates, but their effect on the buoyant force is deemed negligible.

Individual seismic time history accelerations are applied to [ ] nodes of the seismic load model through the LS-DYNA command

                  *BOUNDARY_PRESCRIBED_MOTION_ NODE in three spatial directions.

Seismic motions start at [ ] seconds. The nodes that receive seismic input are highlighted red in Figure 4-31. [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-31 Nodes for Seismic Input [

[ ] seismic acceleration time histories [

                                      ] have a length of approximately 80 seconds ( [
                                                       ] ) in each orthogonal direction. The acceleration time histories in each direction are defined with [                 ] time steps and a time step length of [           ] seconds. [
                               ] A representative set of input accelerations is shown for the center SFP floor node 75634 in Figure 4-32, Figure 4-34, and Figure 4-36 for the x-, y-,

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 and z-direction, respectively. The final analysis time is based on the Arias Intensity measure. The Arias Intensity, E, is calculated as shown in Equation 4-11. The Husid diagram that shows the development of the earthquake intensity over time is defined by: 1 t 2 E o³ H ( t ) = --- a ( )d Equation 4-15 Figure 4-33, Figure 4-35, and Figure 4-37 show the Husid diagrams for the x-, y-, and z-direction, respectively. The times when the motions attain an Arias Intensity of [ ] is noted. The earthquake strong motion duration is defined by the [ ] Arias Intensity. Figure 4-32 [ ] Node 75634, Acceleration Time History, X-Direction [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-33 [ ] Node 75634, Husid Diagram, X-Direction [

Figure 4-34 [ ] Node 75634, Acceleration Time History, Y-Direction [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-35 [ ] Node 75634, Husid Diagram, Y-Direction [

Figure 4-36 [ ] Node 75634, Acceleration Time History, Z-Direction [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-37 [ ] Node 75634, Husid Diagram, Z-Direction [

4.4.2.10 Solution Time Steps Although the WPA Model in LS-DYNA is subjected to acceleration time histories at [ ] intervals, the program automatically calculates the required time steps in each iteration based on the material properties and smallest element size for each element with following equation (for shell elements): L dt = ------------------------------- Equation 4-16 E 2 (1 - v ) where: L = smallest element size E = the modulus of elasticity v = Poissons ratio

                        = density.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The most critical elements with regards to the time step size belong [

                                        ] (Figure 4-38). The controlling time step is [
                                                                                                            ] For stability reasons the scale factor TSSFAC is set to [             ] resulting in a final analysis time step of [                                ] The acceleration time histories contain a [                                                                        ] at the end. In addition, since the strong motion duration of the motions lasts

[ ] (Figure 4-33, Figure 4-35, and Figure 4-37) the vast majority of seismic energy is applied to the WPA Model in that time frame. Therefore, the solution is performed up to [

Figure 4-38 Whole Pool Analysis, Smallest Elements [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.5 Fuel Storage Rack Stress Analysis 4.5.1 Acceptance Criteria The following results from the WPA are checked in order to qualify the design of the FSR in the SFP: Stresses in the FSR members Maximum force acting on the FA grids FSR margin to overturning FSR margin to sliding In addition, friction loads on the SFP floor and maximum FSR-to-FSR impact forces are documented. Per NF-3121 and NF-3251 of Subsection NF to Section III of the ASME B&PV Code (Reference 10.2.b.i), primary membrane and bending stresses are defined as normal stresses. For plate and shell type supports, maximum stress intensity is compared to the allowable stress values. 4.5.1.1 Level A and B Stress Acceptance Criteria The FSR, except for the support legs, is designed using the plate and shell support type acceptance criteria per Subsection NF to Section III of the ASME B&PV Code (Reference 10.2.b.i) for Level A and B loading (Table 4-9). The FSR support legs are designed using the linear support type criteria per Subsection NF as shown in Table 4-9. Per Table NF-3251.2-1 of Subsection NF to Section III of the ASME B&PV Code (Reference 10.2.b.i), allowable primary stress limits for Level B are 1.33 times Level A stress limits. The deadweight analysis uses Level A stress limits and accounts for Level A and Level B Service Limit criteria except the stuck fuel load case, which is evaluated separately using hand calculations. © Copyright 2023 by NuScale Power, LLC 85

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-9 Level A & B Allowable Stress Criteria for Plate and Shell and Linear Type Supports Support Type Stress Type Stress Intensity Limits(1) Reference(2) Membrane Stress Pm 1.0S NF-3251.1 Membrane and Bending Pm + Pb 1.5S NF-3251.1 Plate and Shell Bearing Sy NF-3252.1 Average Primary Shear 0.6S NF-3252.2 Maximum Primary Shear 0.8S NF-3252.2 Tension Ft = min(0.6Sy; 0.5Su) NF-3322.1(a)(1) Shear Fv = min(0.4Sy; 0.33Su) NF-3322.1(b)(1) Linear If K*l / r 120 then Compression(3) Fa = Sy (0.47 - K*l / 444r) NF-3322.1(c)(2) Bending Fb = min(0.6Sy; 0.5Su) NF-3322.1(d)(5)(a) Notes:

1. Pm = Primary membrane stress Pb = Primary bending stress S = Allowable stress from Part D to Section II of the ASME B&PV Code (Reference 10.2.a.iii)

Sy = Yield stress of material Su = Ultimate stress of material Fa = Allowable stress in compression Fb = Allowable bending stress of material Vu = ultimate shear loading K = Effective length factor; used specifically for compression / buckling analysis l = Length of member r = Radius of gyration of member

2. References are from Subsection NF to Section III of the ASME B&PV Code (Reference 10.2.b.i). Stress limits for Class 3 plate and shell type supports shall be in accordance with NF-3250 (Class 2 supports) per NF-3260. Stress Limits for Class 3 linear type supports shall be in accordance with NF-3320 and NF-3340 (Class 1 supports) per NF-3360.

3. Compressive stress may not exceed 2/3 of the critical buckling stress for Service Level B.

4.5.1.2 Level D Stress Acceptance Criteria The stress in the FSR shall also meet the Service Level D accident condition acceptance criteria. The FSR, except the support legs, is designed using the plate and shell support type plastic analysis acceptance criteria per Section F-1340 of Appendix F to Section III of the ASME B&PV Code (Reference 10.2.b.ii) for Level D loading (Table 4-10). The FSR support legs are designed using the linear support type elastic analysis acceptance criteria given in Section F-1334 of Appendix F to Section III of the ASME B&PV Code (Reference 10.2.b.ii) for Level D loading (Table 4-10). © Copyright 2023 by NuScale Power, LLC 86

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-10 Level A & B Allowable Stress Criteria for Plate and Shell and Linear Type Supports Support and Analysis Type Stress Type Stress Intensity Limits Reference(1) Plate and Shell Membrane Stress Pm 0.7Su F-1341.2 Type Supports Membrane and Bending Pm + Pb 0.9Su F-1341.2 Bearing N/A Except Pinned and Bolted Joints F-1341.6 Shear Vu 0.42Su F-1341.2 Plastic System Analysis Compression Must be < 2/3 of the buckling load (or F-1341.8 stress) Linear Type Membrane Stress Pm 0.7Su F-1341.2 Supports Membrane and Bending Pm + Pb 0.9Su F-1341.2 Bearing N/A Except Pinned and Bolted Joints F-1341.6 Shear Vu 0.42Su F-1341.2 Plastic System Analysis Compression Max. compressive load < 2/3 of the F-1341.8 buckling load (or stress) Linear Type Tension Ft = min(1.2Sy; 0.7Su) F-1334.1 Supports Shear Fv = min(0.72Sy; 0.42Su) F-1334.2 Compression where F-1334 if 0 1.0 is applicable - 2 Elastic System 1 - ---- F-1334.3(b)(1) Analysis 4 Fa = Sy ------------------------------------------------------------------------ 2

                                                                                                                                      -3 1.11 + 0.5 + 0.17 - 0.28 Bending if compact per       Fb = fSy                                                                               F-1334.4(b) & (c)

F-1334.4(b) NF-3322.1(d)(2) Axial Compression + Per NF-3322.1(e)(1) Eq (20) to (22)(3) F-1334.5 Bending NF-3322.1(e)(1)

1. Note 1 of the Notes for Table 4-9 contains parameter definitions.

2. References are from Appendix F to Section III of the ASME B&PV Code (Reference 10.2.b.ii). Per F-1342(b), allowable stresses for Plate and Shell type supports for which loads are generated using plastic system methods shall be per F-1341. Per 1344.2, allowable stresses for Linear type supports for which loads are generated using plastic system methods shall be per F.1341.2.

3. Per F-1334, where appropriate the factor used to increase the allowable from NF is the minimum of 2, 1.167Su / Sy if Su > 1.2Sy, or 1.4 if Su 1.2Sy.

4.5.1.3 Load Combinations Load combinations applicable to FSRs are given in Appendix D of the DSRS Section 3.8.4 (Reference 10.1.b) and are listed in Table 4-11. Noting that certain load types specified by the DSRS [ ] are not applicable to the US460 FSRs, the reduced load combinations actually considered in the stress analysis of the US460 FSRs are also listed. The FSR is evaluated for lifting loads when empty. Stresses are checked using hand calculations for the baseplate. Criteria and allowable stresses for lifting are based on ANSI Standard N14.6, Radioactive Materials - Special Lifting Devices © Copyright 2023 by NuScale Power, LLC 87

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 for Shipping Containers Weighing 10,000 Pounds (4,500 kg) or More (Reference 10.15). Appendix D of DSRS Section 3.8.4 (Reference 10.1.b specifies that the D + L + T0

                  + Pf load combination must meet Level B service limits, where Pf is the upward force on the FSRs caused by a postulated stuck FA. This load combination is evaluated using hand calculations.

Appendix D of DSRS Section 3.8.4 (Reference 10.1.b) also specifies that the functionality of the FSRs is demonstrated for the D + L + Fd load combination, where Fd is the force caused by the accidental drop of the heaviest load from the maximum possible height. The heaviest load potentially dropped on the FSRs is determined to be a fuel assembly as described in Section 4.2. This load combination is evaluated separately in the load drop analysis described in Section 4.2. © Copyright 2023 by NuScale Power, LLC 88

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-11 Level A & B Allowable Stress Criteria for Plate and Shell and Linear Type Supports Applicability Load Combination Acceptance Limit D+L D + L + T0 Level A Service Limits D + L + T0 + E Appendix D of DSRS Section 3.8.4 D + L + Ta + E Level B Service Limits (Reference 10.1.b) D + L + T 0 + Pf D + L + Ta + E Level D Service Limits D + L + Fd Demonstrate functional capability of FSR D Level A / B Service Limits D + Pf Level B Service Limits Reduced Load Combinations for FSR Analysis D + E Level D Service Limits D + Fd Demonstrate functional capability of FSR

1. D = Dead load of the FSR and the weight of FAs stored in a fully loaded FSR L = Live load - the weight of the FAs is considered in the deadweight analysis of the FSRs. There are no other live loads to be considered.

To = Thermal effects and loads during startup, normal operating, or shutdown conditions, based on the most critical transient or steady-state condition. [

Ta = Thermal loads under thermal conditions generated by the postulated pipe break accident, pool swell, and subsequent hydrodynamic reaction loads. [

Pf = The maximum uplift force that can be exerted by the fuel bridge on a stuck fuel assembly. This condition is evaluated separately using a hand calculation. E = Dynamic load due to Operating Basis Earthquake (OBE). [

E = Dynamic load due to SSE. Fd = Impact loads due to the heaviest object dropped from the maximum possible height.

2. A thermal stress analysis performed on the FSR using ANSYS (Reference 10.6) shows that under normal conditions, the maximum stress intensity is [ ]

The normal condition thermal stresses combined with the dead load stresses do not exceed the fatigue stress range conservatively considered from Subsection NF to Section III of the ASME B&PV Code, Table NF-3332.4-1 (Reference 10.2.b.i). Also, Subsection NF treats thermal stresses as self-limiting secondary stresses per paragraph NF-3121.11 and as such they need not be considered for primary stress qualification. © Copyright 2023 by NuScale Power, LLC 89

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.5.2 Service Level A/B Stress Results 4.5.2.1 Member Stresses The allowable stress [ ] is S = 20 ksi per Part D to Section II of the ASME B&PV Code (Reference 10.2.a.iii). NF-3212.1 of Subsection NF to Section III of the ASME B&PV Code (Reference 10.2.b.i) defines stress intensity as twice the maximum shear stress. The lower allowable stresses for membrane stresses per Table 4-9 are S = 20 ksi for Service Level A and 1.33S = 26.6 ksi for Service Level B. The Service Level A stress allowable is used with analysis results capturing Service Level B loading. Table 4-12 shows Service Level A stress ratios for the FSR members [

Gravity loads are [ ] are used to derive dead load stresses. Table 4-12 Maximum Design Level A Stresses [ ] Dead Load Shear Stress Intensity Allowable Member Stress Stress Ratio (ksi) (ksi) (psi) [ ] 20 [ ] [ ] 20 [ ] [ ] 20 [ ] [ ] 20 [ ] [ ] 20 [ ] [ ] 20 [ ] [ ] 20 [ ] [ ] 20 [ ] Table 4-14 shows Service Level A stress ratios for the FSR members [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [

Table 4-13 [ ] [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-14 Maximum Level A Stresses [ ] [

4.5.2.2 Weld Stresses The welds between the [

                                                                           ] The welds between the

[

                                                                                  ] The welds in between the [                                                          ]

Per Part D to Section II of the ASME B&PV Code (Reference 10.2.a.iii) the yield strength [ ] is Sy = 28.7 ksi. The governing allowable strength of the welds is determined in accordance with NF-3324.5(2)(-a) in Subsection NF to Section III of the ASME B&PV Code (Reference 10.2.b.i) for shear on the effective throat of partial penetration groove welds: the weld stress limit is equal to the base metal stress limit defined in NF-3321.1: F w,b = 0.4 S y = 11.5 ksi Equation 4-18 For tension on the effective throat NF-3324.5 (2) (-b) and Table NF-3324.5(a)-1 provide the allowables. For the [ ] weld metal the minimum tensile strength is 75 ksi per Part A to Section II of the ASME B&PV Code (Reference 10.2.a.i). The weld metal strength is: F w,s = 0.3 75 = 22.5 ksi Equation 4-19 Table 4-15 shows the weld stress ratios determined by design shear stress over base metal allowable shear stress. Thereby, the maximum shear stresses are obtained through maximum member intensities (Table 4-12 and Table 4-14) and are used to perform the design checks. The maximum shear stresses of the members are scaled by the ratio of base metal thickness over weld size. This is to capture that the stresses in the base metal are transferred into the weld. The weld © Copyright 2023 by NuScale Power, LLC 92

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 stress ratios due to tensile stresses are determined by design stress intensity (two times design shear stress) over the minimum of the weld strength and the allowable base metal stress of S = 20 ksi. Table 4-15 Weld Check, Service Level A [

The welds [

                                                                                       ] the base metal allowables control. Further evaluations of these welds are unnecessary.

A[ ] weld joins the [ ] The weld has an effective throat thickness [ ] The base metal thickness of the [ ] Since the weld metal strength [ ] times stronger than the base metal, the weld is adequate. 4.5.2.3 Support Leg Stresses Layout of the support leg components is shown in Figure 4-39. The [

                                          ] is welded directly to the [
                                                                            ] The [                                  ]

is welded to the [ ] Shear is checked at Sections A-A and B-B as shown in Figure 4-39. Additionally, [

                               ] are checked.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-39 Support Leg Layout [

Linear support type elastic analysis is used for the support leg design. Per Table 4-9 the design tension allowable is Ft = min(0.6Sy; 0.5Su) = 17.22 ksi and the design shear allowable is Fv = min(0.4Sy; 0.33Su) = 11.5 ksi. Maximum Floor Reaction Under One Leg Contact forces between the leg plates and SFP floor are obtained from the LS-DYNA output. Contact force time histories in the Z-direction represent the vertical reaction time history of a leg. Gravity loads [

                                                        ] are used to derive dead load reaction forces.

For each FSR, the vertical forces due to dead load at each of the [ ] legs are determined. Table 4-16 lists the maximum vertical forces due to dead load for each FSR. The maximum vertical leg force in any FSR leg is [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-16 Maximum Leg Forces due to Deadweight Loading Rack Vertical Force (lb) [ ] [ ] [ ] [ ] [ ] [ ] The maximum pressure between the foot and the SFP floor liner due to deadweight loading is calculated as [ ] Foot The foot (Item 15) is evaluated against punching shear but is also treated as a cantilevered equivalent beam. The foot shear stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. The foot bending stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. [ ] The [ ] weld between the [ ] undergoes vertical loading. The vertical downward force is transferred directly via bearing at the contact surfaces between the [ ] Therefore, there is no check required for this weld. [ ] The [ ] transfers the vertical loads [

                                                                           ] The worst condition for this component is when the leg is at its maximum length [
                                                                   ] as shown in Figure 4-39. The [
                             ] compressive stress ratio is calculated as [              ] which is less than 1.0 and therefore acceptable.

The threaded portion of the [ ] transfers the reaction forces at the base of the cantilever as shown in Figure 4-39. Therefore, the thread is checked to ensure that it is sufficient to transfer the loads. The thread is specified as [

                            ] The thread shear stress ratio is calculated as [               ] which is less than 1.0 and therefore acceptable.

The vertical load could cause local bending of the thread, which is analogous to a small edge cantilever, thus the local bending is checked based on a unit width of the circumference of diameter. The thread bending stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. © Copyright 2023 by NuScale Power, LLC 95

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [ ] The lower portion [ ] is loaded in shear due to the vertical loads transmitted through the [ ] (Section A-A in Figure 4-39). The [ ] shear stress ratio at Section A-A is calculated as [ ] which is less than 1.0 and therefore acceptable. The [ ] compressive stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. [ ] The [ ] weld between the [

                                                                                   ] is not subject to the vertical downward load: the vertical downward load is transferred [
                                                               ] via direct bearing between the interface of underside [                                                                ] Therefore, there is no check required for this weld.

[ ] The upper portion [ ] is loaded in shear (Section B-B in Figure 4-39). The [ ] shear stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable 4.5.2.4 Fuel Storage Rack Stresses Due to Stuck Fuel Assembly As specified in Table 4-11, the D + Pf load combination must meet Level B service limits, where Pf is the upward force on the FSRs caused by the force imparted by the fuel handling machine on a postulated stuck FA. The FA is considered to be stuck [ ] for this analysis. Live loads are not applicable for this analysis. The maximum lifting capacity of the crane is [ ]A maximum net force of [ ] is considered to act upwards at the bottom [ ] This creates compression loading [

[ ] [ ] welds [ ] experience direct shear under a Pf loading. The base metal strength is less than the weld metal strength and the base metal thickness is less than the weld thickness and is therefore the controlling allowable value. The [ ] weld stress ratio is calculated as [ ] , which is less than 1.0 and therefore acceptable. © Copyright 2023 by NuScale Power, LLC 96

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [ ] A Pf load creates compression [ ] Although this compression acts between the bottom [

                            ] it is conservatively considered to act along the entire length [
                           ] for this analysis. The [                    ] compression stress ratio is calculated as [             ] which is less than 1.0 and therefore acceptable.

4.5.2.5 Fuel Storage Rack Stresses Due to Lifting Lifting is performed with a tool that lifts the FSR at a minimum of [

                             ] A spreader bar is assumed to be used such that only a vertical load is carried at each location.

Lifting Criteria Per Section 4.2.1 of ANSI N14.6 (Reference 10.15), the load-bearing members of a lifting device shall be capable of lifting three times the weight of the FSR without generating a combined shear stress or maximum tensile stress at any point in excess of the corresponding minimum tensile yield strength of the construction materials. They shall also be capable of lifting five times that weight without exceeding the ultimate tensile strength of the materials. The [

                            ] components loaded during a lift are [                                            ] at the operating temperature of [                     ] As such, the allowable tensile stress for loading associated with the direct weight of the empty FSR is as follows:

S t = min ( S y 3 ;S u 5 ) = 9.57 ksi Equation 4-20 Per NF-3252.1 of Subsection NF to Section III of the ASME B&PV Code (Reference 10.2.b.i), the allowable average shear stress and bearing stress are as follows: S s = 0.6S t = 5.74 ksi Equation 4-21 S b = S y = 28.7 ksi Equation 4-22 where: S s = Allowable average shear stress S b = Allowable bearing stress © Copyright 2023 by NuScale Power, LLC 97

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Lifting Load The empty FSR weight is: [ ] Equation 4-23 Because the empty FSR is not a critical load, the consideration of a single point failure is not necessary. Considering the load to be shared [ ] the weight carried by each part is: [ = ] Equation 4-24 Using a 2.0 dynamic load factor: F d = 2.0 F lift = [ ] Equation 4-25 The [ ] shear stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. The maximum allowable bearing stress [ ] is: b = S b = 28.7 ksi Equation 4-26 Therefore, the minimum required bearing area for a lifting device is: Ab = Fd b = [ ] Equation 4-27 4.5.3 Service Level D Stress Results 4.5.3.1 Member Stresses The true tensile strength [ ] is Su = 96.5 ksi (Section 4.2.1.3). The minimum allowable stress between membrane, and membrane and bending stress limit of 0.7Su = 67.55 ksi is used for the FSR design (Table 4-10). NF-3212.1 of Subsection NF to Section III of the ASME B&PV Code (Reference 10.2.b.i) defines stress intensity as twice the maximum shear stress. Table 4-17 shows Service Level D shear stresses and stress ratios for the FSR members [

                                       ] Figure A-1 through Figure A-8 are shear stress plots for each of these members. The members having the largest shear stresses are shown.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-18 shows Service Level D shear stress stresses and stress ratios [

                           ] Figure A-9 through Figure A-13 are shear stress plots for each of these members. The members having the largest shear stresses are shown.

Table 4-17 Maximum Design Level D Stress Ratios [ ] Rack Stress Allowable Stress Member Figure Intensity [ ] (ksi) Ratio (ksi) [ ] [ ] [ ] [ ] [ ] [ ] [ ]

     © Copyright 2023 by NuScale Power, LLC

[ ] Table 4-18 Maximum Level D Stress Ratios [ ] Rack Stress Combined Scale Allowable Stress Figure Member Intensity Member [ ] Factor (ksi) Ratio (ksi) [ ] [ ] [

[

                                                                                                                                                                    ]
                                                                                                                                                                    ]

[

TR-145417-NP NuScale US460 Fuel Storage Rack Design Topical Report 100 Revision 0

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.5.3.2 Weld Stresses The welds between the [

                                                                              ] The welds between the

[

                                                                                  . ] The welds in between

[ ] The governing allowable strength of the welds is determined in accordance with Table NF-3312.1(b)-1 in Subsection NF to Section III of the ASME B&PV Code (Reference 10.2.b.i), including the minimum stress limit factor of Kv = 2.0. The weld metal strength is: F w,s = 0.3 75 2.0 = 45 ksi Equation 4-28 The base metal strength is: F w,b = 0.4 S y 2.0 = 23 ksi Equation 4-29 Table 4-19 shows the weld stress ratios determined by design shear stress over base metal allowable shear stress. Thereby, the maximum shear stresses are obtained through maximum member intensities (Table 4-17 and Table 4-18) and are used to perform the design checks. The maximum shear stresses of the members are scaled by the ratio of base metal thickness over weld size. This is to capture that the stresses in the base metal are transferred into the weld. The weld interaction ratios due to tensile stresses are determined by design stress intensity (two times design shear stress) over the minimum of the weld strength and the allowable base metal stress of 0.7Su = 67.55 ksi. © Copyright 2023 by NuScale Power, LLC 101

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-19 Weld Check, Service Level D Max Max Min Base Design Design Shear Tensile Stress Shear Weld Metal Shear Stress Member Stress Stress Intensity Stress Size Thickness Stress Intensity Ratio Ratio (ksi) (ksi) (in) (in) (ksi) (ksi) [ ] [ ] [

[

1 The critical design stress of [ ] is determined by the shear stress of [ ] in [ ] Figure 4-40 illustrates the critical members, exhibiting a maximum shear stress range of [ ] which aligns with the allowable strength of the base metal (23[1/8]/0.165/2.32). Elements above this level are shown in black. It is noted that the plot shows single node singularities where the maximum stresses occur. The singularities are away from the weld locations (grid intersections). Given that the elements adjacent to welds remain significantly below the allowable stress, the welds [ ] are considered adequate. The welds between [

                                                                  ] Since the effective throat thickness is equal or larger than the width of the joined components, the base metal allowable values control. Further evaluations of these welds are unnecessary.

A[ ] fillet weld joins the [ ] The weld has an effective throat thickness [ ] The base metal thickness of the flange is [ ] Since the weld metal strength is [ ] times stronger than the base metal, the weld is adequate. © Copyright 2023 by NuScale Power, LLC 102

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-40 Rack 3, Top Inner Grid Shear Stress [

4.5.3.3 Compressive/Buckling Stress Per Table 4-10 members must be limited to 2/3 of their critical buckling load for both plate and shell type and linear type supports. For the FSR, the most significantly loaded members in compression are the [

                           ] Since [                                      ] are short columns, failure due to elastic buckling does not occur. The critical buckling load for the [
                                                                                  ]

These members receive compressive loading from their own self-weight and inertia as well as inertia [ ] Fluid pressure also provides some compression loading. These loads are accounted for by the total reaction loads shown in Section 4.5.3.4. The maximum vertical reaction at a single support leg is [ ] (Table 4-20). Since there are [ ] support legs, two times this load would correspond to a conservative maximum compression load [

                                                  ] This load is less than the critical buckling load [
                                                            . ] Therefore, these members meet the 2/3 criterion for Service Level D and do not buckle under the maximum compressive loading.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-20 Maximum Support Leg Forces Peak Horizontal X Peak Horizontal Y Peak Vertical Peak Lateral Force Rack Force Force Force (lb) (lb) (lb) (lb) [ ] [ ] [ ] [ ] [ ] [ ] 4.5.3.4 Support Leg Stresses Layout of the support leg components is shown in Figure 4-39. [

                                                                                                ] Shear is checked at Sections A-A and B-B as shown in Figure 4-39. Additionally, [
                               ] are checked.

Linear type support elastic analysis is used for the support leg design. Per Table 4-10, the design tension allowable is Ft = min(1.2Sy; 0.7Su) = 34.44 ksi and the design shear allowable is Fv = min(0.72Sy; 0.42Su) = 20.66 ksi. Maximum Floor Reaction Under One Leg Contact forces between the leg plates and SFP floor are obtained from the WPA LS-DYNA output. Contact force time histories in the Z-direction represent the vertical reaction time history of a leg. For each FSR, the peak force in each direction at each of the [ ] legs is determined. Table 4-20 lists the peak vertical forces of a leg plate for each rack. The maximum vertical force in any FSR leg [ ] occurs [ ] The maximum pressure between the foot and the SFP floor liner is calculated as [ ] The maximum lateral force in any FSR leg is [ ] and occurs [ ] Foot The foot (Item 15) is evaluated against punching shear but is also treated as a cantilevered equivalent beam. The foot shear stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. The foot bending stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. © Copyright 2023 by NuScale Power, LLC 104

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [ ] The [ ] weld [ ] undergoes shear loading from the horizontal force and the moment induced by the eccentricity of the horizontal force. The vertical downward force is transferred directly via bearing between the contact surfaces [

                             ] It is recognized that there is an upward force when uplift occurs, and the weld would take on the additional vertical load due to the gravity of the foot components. However, the gravity weights of those components are small thus the load is negligible. The shear stress ratio [
                                        ] is calculated as [           ] which is less than 1.0 and therefore acceptable.

[ ] The [ ] transfers both the vertical and horizontal loads [ ] The worst condition for this component is when the leg is at its maximum length [

                                                                                           ] (Figure 4-39).

The [ ] shear stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. The [ ] compressive stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. The [ ] bending stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. Section F-1334.5 of Appendix F to Section III of the ASME B&PV Code (Reference 10.2.b.ii) states that criteria from equations (20) to (22) of NF-3322.1(e)(1) shall be satisfied but with allowable stresses as defined in F-1334.5. The [ ] combined axial and bending IR (Equation (20) of NF-3322.1(e)(1), (Reference 10.2.b.i)) is calculated as [ ] which is less than 1.0 and therefore acceptable. The [ ] combined axial and bending IR (Equation (21) of NF-3322.1(e)(1), (Reference 10.2.b.i)) is calculated as [ ] which is less than 1.0 and therefore acceptable. The [ ] combined axial and bending interaction ratio (Equation (22) of NF-3322.1(e)(1), (Reference 10.2.b.i)) is calculated as [ ] which is less than 1.0 and therefore acceptable. Thread The threaded portion [ ] transfers the reaction forces at the base of the cantilever as shown in Figure 4-39. Therefore, the thread is checked to ensure that it is sufficient to transfer the loads. The horizontal force reaction is © Copyright 2023 by NuScale Power, LLC 105

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 transferred via shear [ ] therefore, the shear across the reduced area is used here. The [ ] shear stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. The vertical reaction force and the moment reaction from the horizontal load are transferred via vertical shear across the root of the threads. The available length for the shear per pitch is 3/4P, where P is the thread pitch, thus the total available length for resisting shear is 75% of the engaged thread length at the root of the thread. The thread shear stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable The vertical load could cause local bending of the thread, which is analogous to a small edge cantilever, thus the local bending is checked based on a unit width of the circumference diameter. The thread bending stress ratio, including the minimum stress limit factor of 1.5, is calculated as [ ] which is less than 1.0 and therefore acceptable. [ ] The lower portion [ ] is loaded in shear due to the vertical loads [ ] (Section A-A in Figure 4-39). The [ ] shear stress ratio at Section A-A is calculated as [ ] which is less than 1.0 and therefore acceptable. The [ ] shear stress ratio due to horizontal load is calculated as [ ] which is less than 1.0 and therefore acceptable. The [ ] compressive stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. The [ ] bending stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. Section F-1334.5 of Appendix F to Section III of the ASME B&PV Code (Reference 10.2.b.ii) states that criteria from equations (20) to (22) of NF-3322.1(e)(1) shall be satisfied but with allowable stresses as defined in F-1334.5. The [ ] combined axial and bending interaction ratio (Equation (20) of NF-3322.1(e)(1)) is calculated as [ ] which is less than 1.0 and therefore acceptable The [ ] combined axial and bending interaction ratio (Equation (21) of NF-3322.1(e)(1)) is calculated as [ ] which is less than 1.0 and therefore acceptable. The [ ] combined axial and bending interaction ratio (Equation (22) of NF-3322.1(e)(1)) is calculated as [ ] which is less than 1.0 and therefore acceptable. © Copyright 2023 by NuScale Power, LLC 106

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [ ] The [ ] weld [

                                                                                 ] undergoes axial loading from the vertical force, shear loading from the horizontal force, and the moment induced by the eccentricity of the horizontal force. It is recognized that there is an upward force when uplift occurs, and the weld would take on the additional vertical load due to the gravity of the foot components. However, the gravity weights of those components are small thus the load is negligible. The shear stress ratio for the weld [                                                                ] is calculated as

[ ] which is less than 1.0 and therefore acceptable. [ ] The upper portion [ ] is loaded in shear (Section B-B in Figure 4-39). The [ ] shear stress ratio is calculated as [ ] which is less than 1.0 and therefore acceptable. 4.5.4 Fuel Storage Rack Friction Loads Transferred to the Spent Fuel Pool Floor Slab Contact force time histories in X- and Y-directions for the legs of FSRs are extracted from the WPA LS-DYNA output. Lateral friction loads transferred to the SFP floor slab are calculated from the SRSS of the lateral force time histories in the X- and Y-direction. Peak lateral friction forces are determined for each leg. Table 4-20 lists the peak lateral forces from the leg plates (feet) for each FSR. The peak lateral force acting on the SFP liner is [ ] and occurs at [ ] 4.5.5 Rack-to-Rack Contact Forces Rack-to-rack contact (impact) forces are transferred through the [ ] plates. The [ ] impact force time histories are extracted from the WPA LS-DYNA output. Peak contact forces for each FSR from the seismic time history analysis are listed in Table 4-21. Table 4-21 Maximum Rack-to-Rack Impact Forces [ ] Peak Horizontal X Peak Horizontal Y Peak Vertical Force Force Force [ ] (lb) (lb) (lb) [ ] [ ] [ ] [ ] [ ] [ ] [ ] © Copyright 2023 by NuScale Power, LLC 107

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 4.5.6 Maximum Reaction Forces on a Fuel Assembly There are impact forces [ ] through contact of the FA under seismic input motions. Peak lateral force time histories are obtained through SRSS of contact force time histories in the X- and Y- directions. Peak contact forces are listed in Table 4-22. Table 4-22 Maximum Horizontal and Vertical Forces on a Fuel Assembly Peak Horizontal Peak Horizontal Y Peak Lateral Peak Vertical [ ] X Force Force Force Force (lb) (lb) (lb) (lb) [ ] [ ] [ ] [ ] [ ] [ ] The maximum lateral strength of the spacer grids in the US460 17x17 FA at the beginning and end of life are [ ] respectively, at room temperature. The strength of the FA is proportional to the modulus of elasticity of the grid materials [ ] Modulus of elasticities of [

                     ] are [                                                        ] respectively. Since the modulus of elasticity [                           ] reduces slightly with increasing temperature, the FA assembly crushing strength at 100ºF is slightly lower than

[ ] The maximum lateral force for the FA is [ ] (Table 4-22); well below the crushing strength limit [ ] Therefore, FAs is considered qualified for lateral impact loads. 4.5.7 Rack Overturning Margin of Safety Vertical displacements for the FSR foot plates are evaluated and compared against the corresponding SFP floor displacements. The uplift is computed as the difference between the FSR foot plate and the corresponding floor vertical displacement. The maximum uplift computed is less than [ ] and insignificant. Based on the analysis results it is concluded that there is no uplift and hence no risk of overturning. 4.5.8 Rack Sliding Margin of Safety Horizontal (X- and Y-direction) displacements for the leg plates are evaluated and compared against the corresponding SFP floor displacements. Sliding in the X- and Y-direction is computed as the difference between the FSR leg and the corresponding floor displacement (Table 4-23 and Table 4-24). The computed absolute maximum sliding is [ ] and occurs at [ ] The FSR sliding is representatively shown in Figure 4-41 and Figure 4-42 for leg 1 and all FSRs in the X-, and Y-direction, respectively. In general, it is concluded that the FSRs slide due to © Copyright 2023 by NuScale Power, LLC 108

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 seismic loading. [

Table 4-23 Fuel Storage Rack Sliding, X-Direction [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 4-41 Fuel Storage Racks, Leg 1, Sliding, X-Direction [

Table 4-24 Fuel Storage Rack Sliding, Y-Direction [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-24 Fuel Storage Rack Sliding, Y-Direction (Continued) [

Figure 4-42 Fuel Storage Racks, Leg 1, Sliding, Y-Direction [

4.5.9 Whole Pool Analysis Summary The WPA is a nonlinear analysis of the SFP with [ ] FSRs under seismic loading. The SFP is modeled with a cross sectional area of [ ] and a height of water equal to [ ] Each FSR is approximately [

                           ] located at the bottom of the SFP and submerged in water. A nonlinear model of the whole SFP is developed using the LS-DYNA finite element code and the hydrodynamic coupling between the SFP fluid and the FSRs is simulated in a

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 non-linear analysis. The SFP fluid is modeled with ALE elements and the FSRs are modeled with Lagrangian elements. Hydrodynamic coupling between the ALE and Lagrangian elements is established with the LS-DYNA option (*CONSTRAINED _LAGRANGE_IN_SOLID). Seismic time history analysis of the SFP nonlinear model is performed for three orthogonal input acceleration time histories (simultaneously) for a duration of [ ] The time history analysis is performed using a friction coefficient [ ] between the leg plates and the SFP floor. A summary of the maximum stress and interaction ratios calculated for the FSR members and welds is provided in Table 4-25 and Table 4-26. Table 4-25 Fuel Storage Rack Design Stress/Interaction Ratio Summary [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 4-26 Fuel Storage Racks Critical Part and Weld Stress/Interaction Ratio Summary Item Description Design Critical Element Stress/Interaction Ratio [ ] Level A Member Stress [ ] [ ] Level D Member Stress [ ] [ ] Level D Member Stress [ ] [ ] Level D Member Stress [ ] [ ] Level D Member Stress [ ] [ ] Level D Member Stress [ ] [ ] Level D Member Stress [ ] [ ] Level D Member Stress [ ] [ ] Level D Member Stress [ ] [ ] Level D Shear [ ] [ ] Level D Bending + Axial [ ] [ ] Level D Bending + Axial [ ] [ ] Level D Member Stress [ ] [ Level D Member Stress [ ]

[ Level D Member Stress [ ]

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 5.0 Thermal-Hydraulic Analysis The thermal-hydraulic performance of the FSRs in the SFP are assessed for the following conditions: Calculation of the maximum bulk water temperature in the SFP with one or two trains of SFP cooling in operation. Calculation of the time-to-boil in the SFP upon loss of SFP cooling. CFD analysis of the SFP and FSRs to determine the maximum fuel cladding temperature under bounding initial water temperature conditions. 5.1 Maximum Bulk Spent Fuel Pool Temperature The maximum SFP bulk temperature is calculated for two scenarios: with two cooling trains in operation, and upon loss of one cooling train. 5.1.1 Acceptance Criteria The maximum SFP bulk temperature is evaluated against two acceptance criteria: [ ] with two SFP cooling trains in operation; and [ ] with one SFP cooling train in operation. 5.1.2 Methodology Heat exchanger performance data corresponding to the highest shell-side inlet temperature is used to calculate a Coefficient of Performance (CoP) for the heat exchanger. The CoP is then used to predict the temperature difference between the tube side inlet and the shell side inlet that is required to remove the FA decay heat load. 5.1.3 Results The heat removal rate, tube-side inlet temperatures and shell-side inlet temperature from the heat exchanger data sheet are used to calculate the CoP by dividing the heat removal rate by the difference of the tube-side inlet and the shell-side inlet temperature. [

                                                                         ] A conceptual diagram of the SFP heat exchanger is shown in Figure 5-1.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 5-1 Spent Fuel Pool Heat Exchanger Concept Noting that the shell side and tube side flow rates are equal, a formula for the total decay heat load (Q) is derived using the Log Mean Temperature Difference (LMTD) and performing a heat balance on the shell side of the heat exchanger: Q = Constant x ( T 1 - T 3 ) = CoP x T Equation 5-1 where T1 and T3 are defined in Figure 5-1. The total decay heat load in the SFP is [ ] For the case when there are two cooling trains operating, half of the SFP heat is removed by each train and when there is only one cooling train operating all the SFP heat is removed by a single train. In both cases Equation 5-1 is used to calculate the T, which is then used to calculate the bulk SFP temperature. To include the effect of confounding variables like fouling of the tubes, minor losses, etc., the calculated maximum temperature is increased by 5%, which is conservative given that the acceptance criterion is the maximum SFP temperature (i.e., by increasing the calculated temperature, the margin in the acceptance criterion is reduced). The bulk SFP temperature with two trains of SFP cooling in service is [ ] meeting the acceptance criterion of [ ] For one train of SFP cooling in service the maximum bulk pool temperature is [ ] meeting the acceptance criterion of [ ] 5.2 Spent Fuel Pool Time-to-Boil Calculation A calculation is performed to determine the minimum time available to initiate/restore cooling before bulk boiling begins in the SFP upon loss of both cooling trains. © Copyright 2023 by NuScale Power, LLC 115

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 5.2.1 Methodology The amount of heat necessary to raise the water temperature to the boiling point of 212°F is determined considering the mass of water within the SFP, the minimum SFP water level and the amount of water displaced by the FSRs and the FAs in the FSRs. At the initiation of this event, both SFP cooling trains are assumed to be operating, limiting the initial pool temperature to [ ] The time required to heat the water from [ ] to 212°F is calculated using the total FA decay heat load in the SFP. Boiling begins near the surface of the pool, well away from the FAs because the surface of the pool is at atmospheric pressure, with a saturation temperature of 212 ºF. Deeper in the pool the absolute pressure is greater due to the height of the water column above. Given this increased pressure, the saturation temperature at depth is significantly higher than 212°F, preventing the formation of vapor bubbles. This effect is substantial even a few feet below the pool surface. For example, when the bulk pool temperature is 212°F and boiling is occurring at or near the surface, at a depth of 5 feet below the surface the pressure would be approximately 16.77 psia, and the corresponding saturation temperature is 218.7°F. The minimum SFP water level elevation is [ ] but the elevation of the top of the FSRs is at an elevation [ ] so the minimum height of the water column above the fuel assemblies is approximately [ ] The SFP volume considered in the calculation is based on the [ ] SFP water level of [ ], conservatively reducing the amount of water available for heat absorption. [

[ ] Examples of ambient heat losses are heat transfer through the walls of the SFP and evaporation from the pool surface during the heat-up. [

Forced convection and natural circulation within the SFP produces sufficient mixing such that there is no significant variation in the water temperature in different areas of the pool. Therefore, the heat-up of the pool water is determined based on bulk averaged conditions. Per Table A-19E of the Handbook of Hydraulic Resistance (Reference 10.16) the density of liquid water at atmospheric pressure varies from [ ] 3 to 59.84 lbm/ft at 212°F. For the time-to-boil calculation, the density [ ] is used for conservatism: using a lower value for the density lowers the SFP water mass (mass = density*volume), which reduces the amount of heat required to raise the © Copyright 2023 by NuScale Power, LLC 116

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 temperature of the water to boiling temperature (heat required = mass*specific heat*temperature differential) and consequently reduces the time required for initiation of boiling. [

5.2.2 Results The heat (H) required to raise the SFP water temperature from [ ] to boiling temperature is calculated as follows: H = m x C p x T = [ ] Equation 5-2 where: m = mass of water in the SFP = [ ] Cp = specific heat capacity of water in the [ ] to 212°F range [ ] T = 212°F - [ ] Given the SFP decay heat load, Q, of [ ] the time required for bulk boiling to begin following the loss of SFP cooling is approximately [ ]. 5.3 Computational Fluid Dynamics Analysis to Establish Adequate Decay Heat Removal and Natural Circulation Cold water at very high flow rates is pumped down into the SFP from the cooling water inlets at the top of the SFP. The cold water enters the FSRs [

                                                ] heats up as it traverses over the hot FAs and then exits through the top of the racks due to buoyancy driven natural circulation flow. After exiting the racks, the heated fluid must pass through a large volume of cooler fluid, giving up a large amount of energy in the process. The heated water is then pumped out of the SFP through the outlets at the top of the SFP. The cooling water flow rate considered in the CFD analysis is based on a single train of SFP cooling in operation.

Mixing of the heated fluid exiting the racks and the cooler inlet cooling water occurs due to both forced convection and free convection and determines the temperature of the water entering the bottom of the racks. The calculation of the flow patterns that are developed and the resulting temperature entering the racks is best performed with a CFD code. © Copyright 2023 by NuScale Power, LLC 117

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 DSRS Section 9.1.2 (Reference 10.1.c), Part III, Review Procedure 2.I states the SFP design must be analyzed to verify that: The thermal-hydraulic analysis of the flow through the spent fuel racks is adequate for decay heat removal from the spent fuel assemblies during all anticipated operating and accident conditions. Furthermore, the analysis should show adequate natural circulation of the coolant during all anticipated operating conditions, including full core-offloads during refueling, to prevent nucleate boiling for all fuel assemblies. This paragraph in the DSRS is interpreted to require that the temperature at the outer surface of the fuel rod cladding must be less than the local fluid saturation temperature. The CFD analysis determines the maximum cladding temperature indirectly by calculating the maximum water temperature using a CFD model of the SFP, then determining what the cladding temperature would have to be to allow the heat flux necessary to maintain a steady state condition. 5.3.1 Methodology The SFP is analyzed using a CFD model of the SFP water, the cooling system connections, and the FSRs and FAs in the SFP. The CFD model is created in ANSYS DesignModeler and is meshed using the ANSYS Fluent mesher. The calculation of fluid flow patterns and fluid temperatures is performed using ANSYS Fluent 2021 R2 (Reference 10.17). The SFP volume is considered self-contained with no communication with other water sources by way of the weir wall opening, which includes the refueling pool and reactor pool. [

                                                                    ] This analysis technique eliminates the need to determine the specific distribution pattern of the hotter FAs within the FSRs.

Once the maximum water temperature within the FSRs has been determined, a hand calculation is used to determine the film heat transfer coefficient at the outer surface of the fuel rod cladding, and the cladding surface temperature. An axial peaking factor is applied to the heat flux used in the hand calculation to maximize the cladding temperature. The saturation temperature at the top of the FAs is then calculated using the pressure due to the ambient overpressure and the elevation head resulting from the minimum water level elevation for the SFP. © Copyright 2023 by NuScale Power, LLC 118

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 5.3.2 Modeling The model geometry is created in ANSYS DesignModeler. The racks are modeled as multiple regions of continuous bodies that have appropriate porosity and pressure drop coefficients applied to them. Each FSR is sliced into six vertical segments to easily represent the various available flow areas (porosities) that exist. The active fuel region of the fuel assembly is treated as a separate segment so that the decay heat is assigned to the appropriate region. [

[ ] just above the foot of the FSRs extend out from the vertical members of the FSRs. This extended portion [ ] is explicitly modeled and included in the geometry. [ ] The fuel elevator occupies a small volume within the SFP and thus has negligible impact on the thermal analysis. Additionally, the elevator is located away from the FSRs, and the SFP cooling system inlet/outlet pipes and thus has negligible impact on the natural circulation flow within the SFP. [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 5-2 Top View of Computer Aided Design Model Geometry [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 5-3 Isometric View of Computer Aided Design Model Geometry [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 5.3.3 Meshing The CAD model geometry is meshed with the mesher within ANSYS Fluent (Reference 10.17) using polyhedral elements. A surface mesh of the geometry is generated with adequate resolution to capture the features and the face skewness is improved, such that the final surface mesh has a maximum skewness of < 0.5. The volume mesh is generated using polyhedral cells. The turbulent flow within the SFP is resolved using a k- turbulence model with scalable wall functions that does not require a refined mesh in the near wall region (Section 4.18.1 of the ANSYS Fluent Theory Guide (Reference 10.17.a)). Thus, no prisms are generated in the near wall region. The final mesh generated has a minimum orthogonal quality [

                                               ] A bottom view of the mesh is shown in Figure 5-4, which shows that the mesh is refined in the region of interest (i.e., near the heat generating FSRs).

Figure 5-4 Bottom View of Computer Aided Design Model Mesh [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 5.3.4 Computational Fluid Dynamics Model A pressure based, steady state CFD model using a k- turbulence model with the standard wall function and full buoyancy effects options is used to solve for the heat transfer and fluid flow within the SFP. The spent fuel pool water level is modeled at the nominal water level elevation of [ ] because significantly lower water heights only occur during an accident scenario. This additional water at the top of the pool has an insignificant impact on the localized temperatures within the stored fuel assemblies. However, when performing the follow-on hand calculation to determine the maximum cladding temperature and comparing it to the local saturation temperature for water at the top of the FA under design conditions (with cooling available), [

Make-up water added to the SFP is conservatively modeled to be at the exit temperature of the heat exchanger when only a single cooling train is operational. 5.3.4.1 Solver Settings The SIMPLEC algorithm within ANSYS Fluent (Reference 10.17) is used for the CFD analysis. The first order upwind spatial discretization scheme is used for the momentum, energy, and turbulence solutions. The pressure is discretized using the PRESTO! Scheme. The Under-Relaxation Factors (URFs) used for the various solver equations are listed in Table 5-1. Table 5-1 Solver Under-Relaxation Factors Solver Equation Factor Pressure [ ] Density [ ] Body Forces [ ] Momentum [ ] Turbulent Kinetic Energy [ ] Turbulent Dissipation Rate [ ] Turbulent Viscosity [ ] Energy 0.9 to 1 The simulation is run for a total of [ ] iterations. The energy equation URF is varied during the simulation: the simulation is initiated with an energy URF of 1 for the first 500 iterations to ensure quick convergence, which is then reduced to 0.99 for the next 500 iterations and then is finally reduced to 0.9 for the last [ ] iterations to improve stability. © Copyright 2023 by NuScale Power, LLC 123

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 5.3.4.2 Boundary Conditions The Boundary Conditions (BCs) applied to the CFD model are: Mass flow rate BC is applied to the inlets and Intensity and Hydraulic Diameter is selected as the Turbulence specification method. Outflow BC is applied to the outlets, allowing Fluent to determine the flow rates and temperatures such that mass and energy are conserved. [

The walls and floor of the SFP are modeled as adiabatic (i.e., having zero heat flux) and with zero slip. The decay heat is modeled as an energy source term that is applied to the active fuel region in each rack. One storage rack is assigned an energy source that corresponds to every cell in the rack being occupied by an FA from a core offload, 48 hours after shutdown. This is an unrealistic situation because only one single reactor module with 37 FAs is refueled in that period, but this blanket approach eliminates requirements on the placement of recently offloaded FAs. The decay heat produced by an average FA from the previous cycle offloads (the older FAs) is applied to every cell in the other [ ] storage racks. This is also an unrealistic situation, but it allows the model to provide an indication of potential hot spots for the placement of the recent offload. 5.3.4.3 Porous Zone Settings The different segments of the storage racks are modeled as porous media. The porosity is the volume fraction of fluid within the porous region, or the open volume fraction. Since the natural circulation flow within the racks is expected to be primarily upward, the porosity values are calculated based on the available flow area for vertical flow. Horizontal flow is suppressed by setting the pressure loss coefficients for the horizontal directions to arbitrary values that are about three orders of magnitude larger than the coefficients for the upward direction. Except for flow within the storage racks (where the water flow is expected to be in the laminar range), the pressure losses within the SFP are determined by ANSYS Fluent using the k- turbulence model with the standard wall function and full buoyancy effects options. The pressure loss for water flowing vertically through each FA is based on the pressure loss coefficients experimentally determined for the FA and laminar flow pressure loss coefficient correlations from the Handbook of Hydraulic Resistance (Reference 10.16), since flow within the racks is expected to be laminar. The applicable Reynolds number range of the available pressure loss coefficients for the FAs is 7,600 to 500,000 (i.e., in the turbulent range), whereas the expected water flow conditions within the FSRs yield a Reynolds number of less than 2,000. The use of these correlations for this application is © Copyright 2023 by NuScale Power, LLC 124

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 viewed as conservative since the resulting pressure loss coefficients become very large at low Reynolds numbers, and therefore a conservatively high pressure drop is predicted. The pressure loss for each component of the FA is correlated as a function of the local Reynolds number. Once a suitable pressure loss coefficient correlation has been determined, the conversion process consists of plotting the pressure loss as a function of velocity (i.e., Reynolds number) and curve fitting the results with a second order polynomial equation to separate loss factors into their viscous and inertial resistance components. The governing equation of loss factors applied in ANSYS Fluent are shown in Equations 7.1 and 7.2 in the ANSYS Fluent Users Manual (Reference 10.17.b). 5.3.4.4 Material Properties The solids (baseplates, FSR bumpers, and the FSRs) are modeled [

                                      ] The material properties [                             ] are obtained from Section II, Part D, of the ASME B&PV Code (Reference 10.2.a.iii).

The properties of water for a temperature range of 40°F to 270°F are input to ANSYS Fluent, providing the material properties for the liquid portions of the model. These properties are for a pressure of [ ] which corresponds to the pressure at the bottom of the pool with [ ] of water. If the water is borated, the effect of the boration is neglected because the increase in density due to dissolved boron would be uniform throughout the SFP and would not impact local variations in density due to temperature, which provide the primary driving force for natural circulation cooling in this analysis. 5.3.5 Results The results of the CFD analysis are based on a conservative modeling approach that exaggerates the actual heat load expected for the SFP. This exaggeration is intentional, as it provides a simple method of eliminating questions regarding the most limiting location for the core offload FAs. The SFP has a total of [ ] FSRs. The cooling pattern of the FSRs is impacted by the location of the FCO FAs as these FAs have higher decay heat. To bound all possible scenarios, [ ] CFD analysis cases are evaluated with [ ] different heat load configurations based on the location of storage of the FCO FAs. The active FA region of each of the [ ] FSRs are identified with unique names as shown in Figure 5-5. Table 5-2 lists the [ ] analysis cases with their respective locations of FCO FA storage within the FSRs. The table also lists the maximum peak temperatures of the water for each of the cases. Based on these results the maximum fluid temperature of [ ] occurs when the FCO FAs are stored in the FSR designated as [ ] © Copyright 2023 by NuScale Power, LLC 125

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Several parameters, such as the residuals and mass flow imbalance at the SFP cooling system inlets/outlets, are monitored to ensure that the simulations reached adequate convergence. Figure 5-5 Top View of Spent Fuel Pool Showing Unique Active Fuel Assembly Zone Names [

Table 5-2 Analysis Case and Peak Liquid Temperature [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The peak fuel cladding temperature is computed based on the decay heat dissipated from the FA, the peak local water temperature, and the computed film heat transfer coefficient between the fuel rod and the water. The governing equation is: Q Peak = h c x Area x ( T rod - T water ) Equation 5-3 Forced laminar convection correlations are used to estimate the heat transfer coefficient for the fluid within the fuel assemblies. The fluid convective heat transfer coefficient is characterized by the bounding Nusselt number for laminar flow external to longitudinal tubes. Equations 5.317 through 5.321 in the Handbook of Heat Transfer (Reference 10.18) give values of Nusselt number for laminar flow between a square array of cylinders with a uniform heat flux, as is the case for an array of fuel rods. These equations are functions of the rod pitch, the rod diameter, and the local Graetz number (Gzx). The fuel rods are predicted to have [ ] accumulation following operation within the reactor. The heat transfer coefficient is modified to consider the [ ] thickness. To account for the radial peaking, the temperature rise across the fuel assembly is increased by the peaking factor. The calculated temperature of the rod, Trod, is [ ] based on nominal (not minimum) SFP water level. The saturation temperature for water at the top of the FA, taking into consideration the height of the water above the top of the FA, is [ ] Therefore, there is a temperature margin of at least [ ] until the initialization of localized boiling. As such, the FSR design provides sufficient cooling and flow mixing driven by forced and free convection to meet the criterion of maintaining the peak FA rod temperature below the local saturation temperature. © Copyright 2023 by NuScale Power, LLC 127

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 6.0 Criticality Analysis The FSRs provide criticality control to meet the following acceptance criteria of 10 CFR 50.68(b)(4) (Reference 10.3.a) applicable to storage of FAs in the SFP. If credit is taken for soluble boron, the keff of the spent FSRs loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95 at a 95% probability, 95% confidence level if flooded with borated water, and the keff must remain below 1.0 (subcritical), at a 95% probability, 95% confidence level, if flooded with unborated water. The FSRs are designed for storing 17x17 NuFuel-HTP2' fuel assemblies. [

The analysis uses credit for the decrease in reactivity due to depletion and a fuel storage loading pattern to meet the keff limits. Section 6.5.5 provides the loading pattern and burnup requirements. The FSRs utilize neutron absorber plates made of boron carbide-aluminum metal matrix composite (MMC) as the neutron absorber material. This material is ideal for long-term use in chemical, radiation, and thermal environments of wet storage racks. [

6.1 Methodology 6.1.1 Depletion Analysis Fuel depletion calculations are performed using TRITON/ORIGEN-S as controlled by the T-DEPL module of SCALE Version 6.2.4 (Reference 10.8) to provide burnup dependent nuclide concentrations for the criticality calculations. The 252-group ENDF/B-VII.1 library is selected for this calculation. The T-DEPL module allows simplified data input to the functional modules XSProc, NEWT, ORIGEN, and OPUS. XSProc implements problem-dependent temperature interpolation, calculation of Dancoff factors, and resonance self-shielding using Bondaranko factors with full-range intermediate resonance treatment. NEWT provides regional averaged reaction rates for the ORIGEN depletion module. Finally, OPUS provides a means to edit the depletion results. The nuclides credited for irradiated fuel are shown in Table 6-1. This list is taken from NUREG/CR-6665 (Reference 10.19) with additional nuclides as indicated in the table. A subset of these nuclides may partially migrate to the gap between the fuel and the clad. The nuclide concentration from the TRITON/ORIGEN-S calculations are reduced by the gap release fractions. The gap release fractions are the maximum of the values taken from Table 2 (column Gap Release Phase) and Table 3 of Regulatory Guide 1.183 (Reference 10.20), and Table 2.9 of Pacific Northwest National Laboratory (PNNL) PNNL-18212, Revision 1 (Reference 10.21). The nuclides credited, the maximum gap release fraction, and the number density multiplier are shown in Table 6-1. © Copyright 2023 by NuScale Power, LLC 128

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-1 Nuclides Credited in Depletion Calculation NUREG/CR-6665 Nuclide Element Group Release Multiplier [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-1 Nuclides Credited in Depletion Calculation (Continued) NUREG/CR-6665 Nuclide Element Group Release Multiplier [

6.1.2 Criticality Analysis KENO V.a as controlled by the CSAS5 module of SCALE Version 6.2.4 (Reference 10.8) is used to calculate the effective multiplication factor (keff) of the fuel stored in the SFR. The CSAS5 Control module allows simplified data input to the functional modules XSProc and KENO V.a. These modules process the required cross sections and calculate the keff of the system. The 252-group ENDF/B-VII.1 cross section library is selected for this calculation. XSProc implements problem-dependent temperature interpolation, calculation of Dancoff factors, and resonance self-shielding using Bondaranko factors with full-range intermediate resonance treatment. KENO V.a calculates the keff of a three-dimensional system. A sufficiently large number of neutron histories are run for the calculations to ensure the standard deviation is acceptably low, the source convergence tests are passed, and the chi-square test for normality is passed. 6.1.3 Calculation of the Maximum K The equation below is used to evaluate keff at the 95 percent probability and 95 percent confidence level (k95/95) as required by 10 CFR 50.68 (b) (Reference 10.3.a). 2 12 k 95 95 = k eff + k sys - bias m + C § + + + * + k 2 2 2 2 2

                                                      © k                      ¹                 Equation 6-1 sys    bias    tol        tol where:

k eff = the KENO V.a calculated result k sys = summation of k values associated with the variation of system and base case modeling parameters. bias m =bias associated with the calculation methodology compared to benchmarks. The bias is always calculated as a negative number, subtracting a negative number increased k95/95. © Copyright 2023 by NuScale Power, LLC 130

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 C =confidence level multiplier, assumed to be 2 k =standard deviation of the KENO V.a calculated result bias =standard deviation of the methodology bias, bias m sys =standard deviation of k sys k tol =statistical sum of independent manufacturing tolerances, the depletion uncertainty, and the burnup uncertainty tol =standard deviation of k tol Criticality benchmark calculations are performed to establish the values of bias m and bias . These calculations benchmark the ability of the criticality code to predict the reactivity of a system based on comparison to critical experiments. The criticality benchmark calculations and their applicability to the FSR analyses are documented in Appendix B. Criticality calculations are performed to establish the values of k tol , k sys , tol , and sys . 6.2 Assumptions Unless otherwise stated, the following assumptions are common to the analysis of the SFRs. [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [

6.3 Configuration 6.3.1 Fuel Assemblies The spent FSRs are designed to store 17x17 NuFuel-HTP2' fuel assemblies with a maximum nominal enrichment of 4.95 wt.% 235U and a tolerance of 0.05 wt.% 235U. The FA parameters are provided in Table 6-2 and Figure 6-1. The material properties of the fuel assembly components are provided in Table 6-3. © Copyright 2023 by NuScale Power, LLC 132

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-2 Fuel Assembly Specification for Criticality Analysis Description Nominal Value Tolerance Overall Size 8.426 in x 8.426 in Rod Array 17x17 Number of Fuel Rods 264 Number of Guide Tubes 24 Number of Instrument Tubes 1 Instrument Tube Inside Diameter (ID) 11.43 mm [ ] Instrument Tube Outside Diameter (OD) 12.24 mm [ ] Instrument Tube Material Zircaloy-4 Fuel Pellet Theoretical Density 96.5% [ ] Fuel Pellet Diameter 0.3195 in [ ] Fuel Pellet Length 0.400 in [ ] Fuel Pellet Dish Volume (( [ ] }}2(a),(c),ECI [ ] Fuel Pellet Chamfer and Shoulder Volume (( [ ] }}2(a),(c),ECI Fuel Pellet Total Void Fraction (( [ ] }}2(a),(c),ECI Fuel Rod Clad ID 0.326 in [ ] Furl Rod Clad OD 0.374 in [ ] Fuel Active Length 78.74 in [ ] Fuel Rod Pitch 0.496 in [ ] Fuel Rod Clad Material M5 Guide Tube ID 11.43 mm [ ] Guide Tube OD 12.24 mm [ ] Guide Tube Material [ ] HTP Spacer Grid Material [ ] HTP Spacer Grid Mass [ ] HMP Spacer Grid Material [ ] HMP Spacer Grid Mass [ ] HTP & HMP Grid Height 1.75 in Spacer Grid Elevation of Mid-Point from HMP 2.365 in Bottom of Fuel HTP 22.413 in HTP 42.477 in HTP 62.540 in Fuel Assembly Pitch (( [ ] }}2(a),(c),ECI © Copyright 2023 by NuScale Power, LLC 133

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-1 Fuel Pellet Chamfer and Shoulder Dimension (inches) with Tolerances [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-3 Material Properties Material Density Element Nominal M5 [ ][ ] [ ] [ ] [ ] [ ] [ ] Zircaloy 4 (Instrument & Guide Tubes) 0.237 lbm/in3 Tin 1.38 wt.% Iron 0.21 wt.% Chromium 0.1 wt.% Hydrogen 0.0025 wt.% Nitrogen 0.008 wt.% Oxygen 0.13 wt.% Chlorine 0.002 wt.% Carbon 0.015 wt.% Zirconium 98.1525 wt.% Zircaloy 4 (Spacer Grid) 0.237 lbm/in3 Tin 1.33 wt.% Iron 0.21 wt.% Chromium 0.1 wt.% Carbon 0.014 wt.% Lead 0.0130 wt.% Chlorine 0.002 wt.% Sodium 0.002 wt.% Hydrogen 0.0025 wt.% Nitrogen 0.0065 wt.% Oxygen 0.14 wt.% Zirconium 98.18 [ ][ ] [ ] [ ] [ ] [ ] [ ] [ ] [ ] [ ] B4C 2.52 g/cm3 Aluminum 2.6989 g/cm3 Neutron Absorber [ ]mg 10B/cm2 2.6619 g/cm3 Boron 15.3198 wt.% Carbon 4.2539 wt.% Aluminum 80.4263 wt.% © Copyright 2023 by NuScale Power, LLC 135

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 6.3.2 Description of Fuel Storage System The dimensions of the FSR are shown in Table 6-4. [

                                       ] Figure 6-2 provides the layout of the neutron absorber plates for the FSR.

The material compositions of the FSR are shown in Table 6-3. The analysis for the SFRs uses a pool temperature of [ ] and a water density of [ ] slightly higher than the maximum density of water at [ ] due to the compression of water at the depth of the FSR. Table 6-4 Fuel Storage Rack Design Parameters for Criticality Analysis Description Nominal Value Tolerance [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-2 Fuel Storage Rack Neutron Absorber Plate Layout for Revised Fuel Storage Racks [

6.3.3 TRITON/ORIGEN-S Depletion Model The TRITON/ORIGEN-S model uses nominal dimensions as shown in Table 6-2. Section 4.2.3 of NEI 12-16 Revision 4 (Reference 10.10) states that the lattice depletion code be used in a manner that is consistent with nuclear design calculations. Following common practice for core design work, the dimensions and material densities for all components except the fuel pellet are thermally expanded. However, Section 4.1 of NEI 12-16 states that the depletion analysis is to be © Copyright 2023 by NuScale Power, LLC 137

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 performed using nominal fuel geometric dimensions, which could be interpreted as cold dimensions (not thermally expanded). The use of thermally expanded dimensions in the depletion analysis causes the values of keff in the criticality analysis to be decreased by a small amount. A system bias term is established to remove the effects of thermal expansion in the depletion calculations. The fuel model in the KENO V.a criticality models is divided into 25 axial nodes. A separate TRITON/ORIGEN-S case is run for each axial node. ((

                                                 }}2(a),(c)

Table 6-5 (( }}2(a),(c) (( [

                                                                                                     ] }}2(a),(c)

The burnup steps in TRITON/ORIGEN-S are adjusted at each axial node to closely match the burnup dictated by the conservative axial burnup shapes, shown in Table 6-6, times the assembly average burnup. The assembly average burnup is selected as integer values from 0 to 47 GWD/MTU. This allows nuclide number © Copyright 2023 by NuScale Power, LLC 138

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 densities to be input to the KENO V.a criticality model for depletions of 0, 1, 2, ... 47 GWD/MTU without interpolation. The maximum depletion step is no more than 0.25 GWD/MTU. 6.3.4 KENO-V.a Criticality Model The FSR is modeled in KENO V.a with the nominal FA dimensions shown in Table 6-2 and the nominal rack dimensions shown in Table 6-4. The FA in the KENO V.a model is divided into 25 nodes of equal height. ((

                                                                                                }}2(a),(c)

The axial distribution of burnup for irradiated FAs in the KENO V.a model is either flat (uniform) or taken from a set of conservatively calculated axial burnup distributions. The conservative axial burnup distributions are calculated based on a series of specific core calculations that begin with an initial cycle loaded with new fuel and end with an equilibrium cycle. For all assemblies the burnup distribution is normalized and grouped by average exposure at intervals of 5 GWD/MTU. For each group the minimum normalized value is chosen for the top and bottom 5 nodes. For the middle 15 nodes the average value is chosen. The resulting distribution is not renormalized, resulting in average normalized values below 1.0. The conservative axial burnup shapes are shown in Table 6-6. The NuScale Power Modules operate with all control rods fully withdrawn, with the bottom of the control assembly approximately [ ] above the active fuel column. The control rods have negligible impact on the axial burnup distributions. Table 6-6 Conservative Composite Axial Burnup Shapes (( [

                                                                                                     ] }}2(a)(c)

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-6 Conservative Composite Axial Burnup Shapes (Continued) (( [

                                                                                                      ] }}2(a)(c)

The KENO V.a model of a single storage cell location is shown in Figure 6-3 with corner details shown in Figure 6-4. The single rack location model is used for sensitivity studies, uncertainty analysis, and system bias terms. The whole pool is modeled as a collection of [ ] individual FSRs, arranged as shown in Figure 3-1. The whole pool model conservatively places the racks with [ ] adjacent FSRs nearly touching, with a the nominal spacing to the pool walls. The whole pool model is used for uncertainty and bias analyses to determine the burnup versus enrichment limits and to analyze accident conditions. The KENO V.a model of a single FSR from the whole pool is shown in Figure 6-5. © Copyright 2023 by NuScale Power, LLC 140

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-3 KENO-V.a Model of a Single Fuel Storage Rack Location [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-4 KENO-V.a Model of a Single Fuel Storage Rack Location Showing Corner Details [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-5 KENO-V.a Model of a Single Rack from the Whole Pool Model [

6.4 Initial Conditions, Boundary Conditions, and Limitations The specified operating temperature for the SFP is [

                     ] The criticality analysis is performed at [        ] using the ability of the SCALE module XSProc to internally interpolate cross sections on temperature. The moderator density is slightly more than [                 ] to account for the slight compression of water at the depth of the FSRs.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 6.5 Analysis, Evaluation, and Data 6.5.1 Fuel Storage System Tolerance Analysis Uncertainty factors are evaluated with no soluble boron and with soluble boron at the minimum concentration. Uncertainty factors for manufacturing tolerances that are related to the FA or to the storage cell are determined with a KENO V.a model of a single storage cell location. Uncertainty factors for manufacturing tolerances that are related to an FSR, or the array of FSRs are determined with a KENO V.a model of the whole pool. The uncertainty factors that are burnup dependent are determined with the following methodology. The KENO V.a model of a single storage cell location is run at the burnup points for which the conservative axial burnup shapes are defined, shown in Table 6-6. The KENO V.a model is run with inputs unperturbed except with the parameter of interest perturbed with the specified tolerance factor. Both unperturbed and perturbed cases are performed with a uniform axial burnup distribution and with the conservative axial burnup shapes from Table 6-6. For both the perturbed and unperturbed cases, the values of keff are curve fit to a cubic polynomial against the values of burnup. The uncertainty factor (ktol) is then calculated as the difference between the curve fit value of keff at intervals of 1 GWD/MTU. The standard deviation of the uncertainty factor (tol) conservatively accounts for the maximum standard deviation (k) from both sets of KENO V.a runs and the standard deviation of the residuals from both curve fits. At each burnup point the maximum value is chosen from the uniform axial burnup data and the conservative axial burnup shape data. The standard deviation associated with the maximum values is chosen. This interpolation scheme is used to avoid the ambiguity of which axial burnup shape to use for burnup points between the points where the conservative axial burnup shapes are defined. 6.5.1.1 Uncertainty Factors Dependent on Enrichment and Not Burnup Uncertainty factors for the pin pitch and the clad OD tolerance factors have little variation with burnup. This conclusion is extended to other manufacturing tolerances whose perturbations make small changes to the moderator to fuel ratio. Thus, the uncertainty factors for the ID and OD of the instrument and guide tubes, the storage cell inside width, the storage cell pitch, and the absorber plate width are characterized as a function of enrichment and boron concentration. The manufacturing tolerance for the ID and OD of the instrument and guide tubes are combined into one uncertainty factor by simultaneously increasing the ID and decreasing the OD to maximize the change in moderator to fuel ratio. The uncertainty factors are computed only at zero burnup. The uncertainty factor is the difference between the KENO V.a keff of the unperturbed and the perturbed case, and the standard deviation is the statistical combination of the KENO V.a standard deviation of keff, k. © Copyright 2023 by NuScale Power, LLC 144

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The uncertainty factors for the individual manufacturing tolerances are shown in Table 6-7 and Table 6-8, while the statistically combined uncertainty factors are shown in Table 6-9 and Table 6-10. The uncertainty factors for the Guide and Instrument Tubes (GT ITs), the neutron absorber plate width, and the cell pitch are small and exhibit a significant amount of scatter. The maximum value of these three uncertainty terms are used instead of the enrichment dependent values to compute the statistically combined uncertainty factors. Table 6-7 Manufacturing Tolerance Uncertainty Factors versus Enrichment, 0 ppm Boron [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-8 Manufacturing Tolerance Uncertainty Factors versus Enrichment, 1450 ppm Boron [

Table 6-9 Combined Uncertainty Factors versus Enrichment, 0 ppm Boron [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-10 Combined Uncertainty Factors versus Enrichment, 1450 ppm Boron [

6.5.1.2 Uncertainty Factors Dependent on Enrichment and Burnup The uncertainty factors dependent on enrichment and burnup include the following: Fuel pellet loading Pellet enrichment Reactivity change due to depletion Burnup These uncertainty factors are evaluated with the burnup interpolation methodology discussed in Section 6.5.1. Uncertainty for Fuel Pellet Loading The fuel pellet loading factor accounts for the manufacturing tolerances that affect the amount of fuel in an assembly, including manufacturing tolerances for the [

                                                        ] as shown in Figure 6-1. The [                         ]

manufacturing tolerance is assumed to be independent of the [ ] manufacturing tolerances. The nominal fuel void fraction is [ ] the total of the [

                                                                                            ] The minimum volume of the [                             ] is computed with the worst combination of dimensions and manufacturing tolerances from Figure 6-1 along with the

[ ] and manufacturing tolerances from Table 6-2. The minimum volumes of the [ ] gives a minimum fuel void fraction of [ ] Statistically combining the minimum fuel void fraction with the fuel density tolerance of [ ] on a relative basis, gives a fuel loading tolerance of [ ] The uncertainty factor for fuel loading tolerance is evaluated by two different methods. In the first method, the fuel loading is increased by the fuel loading tolerance in the TRITON/ORIGEN-S cases that then provides fuel number © Copyright 2023 by NuScale Power, LLC 147

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 densities to the KENO V.a model of a single storage cell location. In the second method the KENO V.a model of a single storage cell location uses fuel number densities from a nominal TRITON/ORIGEN-S cases that are increased by the fuel loading tolerance. In both methods the KENO V.a results from the perturbed cases are compared to KENO V.a nominal results. Fuel Enrichment Uncertainty Factor The uncertainty factor for fuel enrichment is evaluated by changing the enrichment in the TRITON/ORIGEN-S cases by 0.05 wt.% 235U. These cases then provide fuel number densities for the KENO V.a model of a single storage cell location. The uncertainty factor is computed by comparing the results from the perturbed KENO V.a cases to nominal results. Depletion Uncertainty Factor A depletion uncertainty of 5% of the reactivity decrement due to depletion as allowed by Section 4.2.3 of NEI 12-16 Revision 4 (Reference 10.10). The KENO V.a model of a single storage cell location is run with nominal nuclide concentrations from TRITON/ORIGEN-S cases. The reactivity uncertainty is computed as 5% of the keff difference from zero burnup to the burnup value of interest, using the same curve fit to determine both values of keff. The standard deviation of the uncertainty is conservatively taken as the statistical combination of the maximum standard deviation of the KENO V.a cases and the standard deviation of the curve fit residuals. Both standard deviations are taken twice in the statistical combination. Reactor Record Burnup Uncertainty Factor The burnup uncertainty factor is taken as the reactivity equivalent of 5% of the stated burnup value as discussed in Section 5.1.5 of NEI 12-16 Revision 4 (Reference 10.10). The burnup uncertainty is evaluated as the difference in keff when evaluated at 0.975 of the burnup value of interest and at 1.025 of the burnup value. The same curve fit is used to determine both values of keff. The standard deviation of the uncertainty is conservatively taken as the statistical combination of the maximum standard deviation of the KENO V.a cases and the standard deviation of the curve fit residuals. Both these standard deviations are taken twice in the statistical combination. Combined Uncertainty Factor versus Enrichment and Burnup The combined uncertainty factor versus enrichment and burnup is the statistical combination of the uncertainty factors for the four parameters discussed above. The standard deviation for the total uncertainty is also computed as the statistical combination of the standard deviation of the four parameters discussed above. The total uncertainty factor is shown in Table 6-11 and Table 6-12, and the standard deviations are shown in Table 6-13 and Table 6-14. © Copyright 2023 by NuScale Power, LLC 148

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-11 Combined Uncertainty Factor versus Enrichment and Burnup with No Soluble Boron [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-11 Combined Uncertainty Factor versus Enrichment and Burnup with No Soluble Boron (Continued) [

Table 6-12 Combined Uncertainty Factor versus Enrichment and Burnup with 1450 ppm Soluble Boron [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-12 Combined Uncertainty Factor versus Enrichment and Burnup with 1450 ppm Soluble Boron (Continued) [

Table 6-13 Standard Deviation of Combined Uncertainty Factor versus Enrichment and Burnup with No Soluble Boron [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-13 Standard Deviation of Combined Uncertainty Factor versus Enrichment and Burnup with No Soluble Boron (Continued) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-14 Standard Deviation of Combined Uncertainty Factor versus Enrichment and Burnup with 1450 ppm Soluble Boron [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-14 Standard Deviation of Combined Uncertainty Factor versus Enrichment and Burnup with 1450 ppm Soluble Boron (Continued) [

6.5.1.3 Uncertainty Factors Dependent on Fuel Storage Rack Configuration Uncertainty factors for fuel stack height and fuel stack axial position are determined with the KENO V.a model of the whole pool. This model allows the neutronic interaction of the fuel with the structural elements outside of the storage cell array and allows the neutronic interaction between FSRs. The height of the fuel above the base plate in the whole pool FSR model is determined by the height of the fuel above the bottom of the FA lower tie plate. The elevation of the structural elements outside of the storage cell array is tied to the top of the base plate. Changing the length of the fuel column or changing the elevation of the fuel column merely changes the position of the structural elements with respect to the fuel column. The storage cell walls and absorber plates are always exactly even with the top and bottom of the fuel column. The uncertainty factors are computed with the whole-pool model with the fuel loading pattern shown in Figure 6-6 and the assembly eccentric positions as shown in Figure 6-7. This is the loading pattern and eccentric positions that are used to set the burnup versus enrichment limits in Section 6.5.5. The fuel enrichments and burnup values at which the analysis is performed are shown in Table 6-15, along with the burnup limits from Table 6-35. Small discrepancies in the burnup value used in the uncertainty analysis and the burnup limits do not cause a significant impact to the results. [

                                                      ] For each uncertainty factor the maximum value is taken over the enrichment and burnup values. The results are shown in Table 6-16.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-6 Nominal Fuel Loading Map (Map Option 5) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-7 Assembly Eccentric Position Map (Scenario 3) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-15 Final Burnup versus Enrichment Limit for Zone 1 Enrichment Burnup Limit Analysis Burnup (wt.% 235U) (GWD/MTU) (GWD/MTU) [

Table 6-16 Uncertainty Factors and Standard Deviation from Whole Pool Model Tolerance Boron Concentration Uncertainty Standard Deviation Stack Height 0 ppm [ ] Stack Height 1450 ppm [ ] Stack Axial Position 0 ppm [ ] Stack Axial Position 1450 ppm [ ] 6.5.2 Fuel Assembly Physical Changes with Depletion NEI 12-16 Section 4.2.2 (Reference 10.10) requires consideration of changes in the fuel rod (clad creep and fuel densification and swelling) and changes in the spacer grid in the criticality analysis. 6.5.2.1 Spacer Grid Growth The spacer grid outer envelope is expected to grow with depletion as shown in Table 6-17 as (%L/L). The projected growth is based on post irradiation examination of spacer grids of the same design that operated in a reactor with similar operating conditions. The growth is shown as an upper tolerance level. The spacer grid envelope growth is assumed to apply to the pin pitch and the pin pitch growth, as shown in Table 6-17. [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-17 Spacer Grid and Pin Pitch Growth Burnup Limit Growth Delta Pin Pitch (GWD/MTU) (%L/L) (in) [

6.5.2.2 Fuel Pellet Densification and Swelling and Clad Creep Appendix B of NEI 12-16 (Reference 10.10) provides a generic analysis that concludes that fuel pellet changes and clad changes, when considered holistically, produce a small negative reactivity change for fuel in a KENO V.a single cell FSR model and therefore is ignored. The NuScale-HTP2 fuel design is analyzed using the Framatome COPERNIC (Reference 10.22) fuel performance program for nominal conditions. The COPERNIC results are shown in Figure 6-8 for pellet density, Figure 6-10 for pellet diameter, Figure 6-12 for clad OD, and Figure 6-14 for clad thickness. For convenience the equivalent results from NEI Appendix B are shown in Figure 6-9, Figure 6-11, Figure 6-13, and Figure 6-15. [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-8 Fuel Pellet Density versus Burnup [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-10 Fuel Pellet Diameter versus Burnup [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-12 Clad Outside Diameter versus Burnup [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-14 Clad Thickness versus Burnup [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-15 NEI 12-16 Clad Thickness versus Burnup 6.5.3 System Bias for Modeling Parameters 6.5.3.1 System Bias for Moderator Inside the Instrument and Guide Tubes [

The system bias term is calculated based on a single FSR model for each enrichment, with both uniform axial burnup shapes and with the conservative axial burnup shapes shown in Table 6-6. Additionally, calculations are performed both without soluble boron and with the minimum soluble boron concentration. The fuel compositions are taken from nominal TRITON/ORIGEN-S depletion cases. For each boron concentration the system bias term is taken as the maximum value over all enrichment and burnup values, while the standard deviation is associated with the maximum value. The maximum value over all enrichments and burnups is applied as a bias. The results are shown in Table 6-18. © Copyright 2023 by NuScale Power, LLC 166

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-18 System Bias for Moderator Inside the Instrument and Guide Tubes Boron Concentration System Bias Standard Deviation 0 ppm [ ] 1450 ppm [ ] 6.5.3.2 System Bias for Thermal Expansion in TRITON/ORIGEN-S Depletion As discussed in Section 6.3.4, the effects of thermally expanding the dimensions and material densities for the TRITON/ORIGEN-S cases is addressed by applying a system bias term. KENO V.a cases for the single FSR location are performed for each enrichment with: Uniform axial burnup shapes Conservative axially distributed burnup shapes (Table 6-6) Without soluble boron The minimum soluble boron concentration. The fuel compositions are calculated using both nominal TRITON/ORIGEN-S depletion cases that include thermal expansion and TRITON-ORIGEN-S cases without thermal expansion. The system bias is computed by comparing the results from the nominal KENO V.a cases to KENO V.a cases using TRITON/ORIGEN-S data without the thermal expansion. The reactivity effect is determined using the depletion interpolation methodology discussed in Section 6.5.1. The maximum bias factor with its associated standard deviation is chosen for each burnup step. The bias factors are shown in Table 6-19 and Table 6-21 while the standard deviations are shown in Table 6-20 and Table 6-22. Table 6-19 Thermal Expansion Bias with No Soluble Boron [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-19 Thermal Expansion Bias with No Soluble Boron (Continued) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-20 Standard Deviation of Thermal Expansion Bias with No Soluble Boron [

Table 6-21 Thermal Expansion Bias with 1450 ppm Soluble Boron [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-21 Thermal Expansion Bias with 1450 ppm Soluble Boron (Continued) [

Table 6-22 Standard Deviation of Thermal Expansion Bias with 1450 ppm Soluble Boron [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-22 Standard Deviation of Thermal Expansion Bias with 1450 ppm Soluble Boron (Continued) [

6.5.3.3 System Bias for Damaged Fuel Cladding A damaged FA is assumed to have lost cladding integrity such that the fuel pellet to clad gap is completely flooded with moderator for all fuel pins. The whole pool model is used to establish a system bias term for this effect. The reactivity effect of the flooded gap is determined both without soluble boron and with soluble © Copyright 2023 by NuScale Power, LLC 172

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 boron at the minimum concentration. The new fuel assemblies are modeled at the maximum reactivity (4.95 wt.% 235U enrichment) and with the pellet-clad gap flooded with moderator in all the fuel pins. Depleted fuel is assumed to have enrichment and burnup values as shown in Table 6-15. The results with damaged fuel cladding are compared to nominal cases that are otherwise identical. The resulting system bias for cladding damage is shown in Table 6-23. Table 6-23 System Bias for Cladding Damage Enrichment Assembly Average Soluble Boron Bias for Clad Standard Burnup (ppm) (wt.% 235U) (GWD/MTU) Damage Deviation 0 [ ] 0 [ ] 0 [ ] 0 [ ] 0 [ ] 0 [ ] 0 [ ] 1450 [ ] 1450 [ ] 1450 [ ] 1450 [ ] 1450 [ ] 1450 [ ] 1450 [ ] 6.5.3.4 System Bias for Pin Pitch Growth A bias for pin pitch growth is determined using the whole pool model. The reactivity effect of pin pitch growth is determined both without soluble boron and with soluble boron at the minimum concentration. The new fuel assemblies are modeled at the maximum reactivity (4.95 wt.% 235U enrichment). Depleted fuel is assumed to have enrichment and burnup values as shown in Table 6-15. Evaluations for enrichments of [ ] are not included because [

                            ] Pin pitch is increased in the depleted fuel assemblies by the values from Table 6-17. [
                                                     ] The results with increased pin pitch are compared to nominal cases that are otherwise identical. The resulting bias for pin pitch growth is shown in Table 6-24.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-24 System Bias for Pin Pitch Growth Enrichment Assembly Average Soluble Boron Bias for Grid Standard Burnup (ppm) (wt.% 235U) (GWD/MTU) Growth Deviation 0 [ ] 0 [ ] 0 [ ] 0 [ ] 0 [ ] 1450 [ ] 1450 [ ] 1450 [ ] 1450 [ ] 1450 [ ] 6.5.4 Fuel Storage Configurations and Assembly Eccentric Positioning Scenarios The allowable fuel storage configurations for the FSRs are shown in Figure 6-6. Any assembly with a nominal enrichment up to 4.95 wt.% 235U may be placed in a Zone 2 location. An assembly must meet the enrichment and burnup requirements of Table 6-15 prior to being placed in a Zone 1 location. Fuel loading maps for a single misloaded FA are shown in Figure 6-16 through Figure 6-22. A fuel loading map to analyze multiple misloaded new FAs is shown in Figure 6-23. A fuel loading map to analyze FSR motion due to a seismic event is shown in Figure 6-24. The analysis evaluates 5 different scenarios of FA eccentric positioning within individual storage locations. Sensitivity studies are performed to determine the most reactive eccentric positioning scenario. [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [

Figure 6-16 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 6) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-17 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 7) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-18 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 8) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-19 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 9) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-20 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 10) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-21 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 11) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-22 Fuel Loading Map for a Single Misloaded Fuel Assembly (Map Option 12) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-23 Fuel Loading Map for Multiple Misloaded Fuel Assemblies (Map Option 13) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-24 Fuel Loading Map for Seismic Induced Rack Motion (Map Option 14) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-25 Assembly Eccentric Positioning Map (Scenario 1) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-26 Assembly Eccentric Positioning Map (Scenario 2) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-27 Assembly Eccentric Positioning Map (Scenario 4) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-28 Assembly Eccentric Positioning Map (Scenario 5) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-25 Sensitivity Study of Eccentric Positioning for the Nominal Fuel Loading Map Eccentric Boron Positioning Concentration keff k k95/95 Description Delta k Scenario (ppm) 0 [ ] 1 1450 [ ] 0 [ ] Option (2-1) [ ] 2 1450 [ ] Option (2-1) [ ] 0 [ ] Option (3-1) [ ] 3 1450 [ ] Option (3-1) [ ] 0 [ ] Option (4-1) [ ] 4 1450 [ ] Option (4-1) [ ] 6.5.5 Burnup and Enrichment Limits for Zone 1 The whole pool model is used to determine the burnup versus enrichment curve for Zone 1 fuel. The nominal fuel loading map (Option 5) shown in Figure 6-6 is used. A sensitivity study determines the limiting eccentric positioning scenario for the nominal fuel loading map, as discussed in Section 6.5.4. The Zone 1 fuel assembly has a [ ]. The model uses the conservative axial burnup shapes from Table 6-6. The fuel composition is taken from nominal TRITON/ORIGEN-S depletion cases. The Zone 2 fuel assembly has a 4.95 wt.% 235U enrichment with no burnup. The results shown in Table 6-25 indicate that eccentric positioning scenario 3 (Figure 6-7) is the most reactive. For each enrichment, KENO V.a cases are performed with the model of the whole pool, zero soluble boron, the nominal fuel loading map (Figure 6-6), and eccentric positioning scenario 3 (Figure 6-7). For these cases, Zone 2 is new fuel enriched to 4.95 wt.% 235U and Zone 1 fuel varies burnup over a range that ensures that the chosen burnup limit has several burnup points higher and lower. The exception is for enrichment of [ ] wt.% 235U, which does not have enough reactivity at zero burnup to exceed the final K95/95 analytical value. The cases examine a uniform burnup distribution and the conservative axial burnup shapes from Table 6-6. For burnups between the values corresponding to axial burnup shapes, a case is performed for both axial shapes to bound the selected burnup value. The data are plotted for each enrichment in Figure 6-29 through Figure 6-35. The fuel compositions for depleted fuel come from TRITON/ORIGEN-S cases that include a neutron source assembly. This removes any restriction on the number or location of neutron source assemblies during reactor operations. © Copyright 2023 by NuScale Power, LLC 188

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-29 k95/95 Versus Burnup for Fuel Enriched to 4.95 wt.% 235U [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-30 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-31 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-32 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-33 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-34 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-35 k95/95 Versus Burnup for Fuel Enriched to [ ] wt.% 235U [

[

Determining the burnup limit versus enrichment is a two-step process, with the goal of achieving a burnup limit of [ ] 235U and having the same margin of k95/95 for all enrichments. For each enrichment except [ ] wt.% 235U a quadratic curve fit of k95/95 versus burnup is computed. The constant term of the curve fit is adjusted so that the curve fit value of k95/95 is always greater than or equal to the individual values of k95/95. A © Copyright 2023 by NuScale Power, LLC 195

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 preliminary burnup limit is computed from the adjusted curve fit coefficients and a target value of k95/95. For [ ] wt.% 235U enrichment, the preliminary burnup limit is linearly extrapolated from the lowest two burnup points. This extrapolation yields an unreal negative burnup limit. A quadratic curve fit of the preliminary burnup limits versus enrichment is computed. This curve fit includes the negative burnup limit for [ ] wt.% 235U enrichment. The constant value of the curve fit is adjusted so that the curve fit value of burnup is always greater than or equal to the individual values of burnup. The target value of k95/95, used to determine the preliminary burnup limit for each enrichment, is adjusted so that the burnup limit for [ ] wt.% 235U enrichment from the adjusted curve fit of burnup versus enrichment is slightly less than the desired value of [

                            ]. The target value of k95/95 is [               ] giving a k95/95 margin of

[ ] for nominal conditions and no soluble boron. The burnup versus enrichment limits for Zone 1 are determined from the following formula: 2 BU A x E + B x E + C Equation 6-2 Where: Burnup is in units of GWD/MTU E is nominal assembly average enrichment in units of wt.% 235U [ ] [ ] The burnup versus enrichment limits are shown in Figure 6-36 and Table 6-15. © Copyright 2023 by NuScale Power, LLC 196

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-36 Final Burnup versus Enrichment Limit for Zone 1 [

Thus, fuel enriched to [ ] wt.% 235U or less may be placed anywhere in the FSRs without limitation. 6.5.6 Damaged Fuel Assembly The SFRs are designed to store up to [ ] damaged FAs. For this analysis a damaged FA is assumed to be discharged to the spent fuel pool before achieving a burnup that allows storage in Zone 1. Damaged FAs are conservatively assumed to be new FAs and must be placed in Zone 2 locations. 6.5.7 Assembly Dropped on Top of Rack An FA dropped on the top of the rack deforms the upper grid structure by less than [ ] The distance from the assembly on top of the rack to an assembly correctly stored in the rack is approximately [ ] accounting for deformation. Section 6.3.1 of NEI 12-16 revision 4 states that a separation of 12 inches is sufficient to prevent neutronic coupling to the dropped assembly. © Copyright 2023 by NuScale Power, LLC 197

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [

6.5.8 Burnup and Enrichment Selection for Accident Analysis The whole pool model is used to analyze accident conditions in the following sections: Assembly Dropped Outside of the FSR Single Misloaded Fuel Assembly Multiple Misloaded Fuel Assemblies Spent Fuel Pool Seismic Event For accident analyses, except the multiple misload analysis, the whole pool model is used with Zone 2 fuel enriched to 4.95 wt.% 235U with no burnup, and with Zone 1 fuel enriched to [ ]. This combination of enrichment and burnup for Zone 1 is taken from Table 6-15. The methodology employed by the analysis in Section 6.5.5 ensures that each combination of burnup and enrichment in Table 6-15 has the same margin of k95/95. Thus, only one combination of enrichment and burnup is analyzed. The accident analysis is only performed with a uniform axial burnup distribution as Section 6.5.5 showed that this is conservative. 6.5.9 Assembly Located Outside of the Fuel Storage Rack FAs are located next to the middle FSR on the right-hand side of the FSR layout shown in Figure 3-1. Four locations are analyzed as shown in Figure 6-37 through Figure 6-40, which show the entirety of one FSR and the dropped FA. The FA is located as close as possible to the FSR [

                             ] This is shown in Figure 6-41. Mislocated FAs are held at the same elevation as the FAs stored in the rack even though the FA would be located axially several inches lower on the SFP floor.

The fuel in the FSRs is loaded according to the nominal load map as shown in Figure 6-6, using eccentric positioning scenario 5, shown in Figure 6-28. This option is like option 3, shown in Figure 6-7, which is used to set the burnup versus enrichment limits discussed in Section 6.5.5. The difference is that scenario 5 moves the assemblies in the FSR closer to the mislocated assembly to increase neutron coupling. The location on the mislocated fuel assemblies outside the FSR is shown in Figure 6-37 through Figure 6-40. © Copyright 2023 by NuScale Power, LLC 198

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-37 Mislocated Fuel Assembly Location 1 [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-38 Mislocated Fuel Assembly Location 2 [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-39 Mislocated Fuel Assembly Location 3 [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-40 Mislocated Fuel Assembly Location 4 [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-41 Mislocated Fuel Assembly Vertical View [

The fuel compositions are calculated with TRITON/ORIGEN-S including the presence of a neutron source assembly. The analysis is performed with soluble boron at the minimum concentration. A mislocated FA is an off-normal condition, therefore the minimum soluble boron concentration is credited. The results are shown in Table 6-26. [

Table 6-26 Dropped Fuel Assembly Results Mislocated Boron Delta k for Fuel keff k k95/95 Concentration Mislocated Std Dev Assembly (ppm) Assembly Location 1 1450 [ ] 2 1450 [ ] 3 1450 [ ] 4 1450 [ ] none 1450 [ ] © Copyright 2023 by NuScale Power, LLC 203

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 6.5.10 Single Misloaded Assembly A single misloaded FA is analyzed in one of seven potential locations as shown in Figure 6-16 through Figure 6-22. The Zone 1 fuel is modeled with [

                                                   ] The misloaded assembly analysis is performed with the uniform axial burnup distribution as this burnup distribution was shown to be conservative in Section 6.5.5. The Zone 2 fuel as well as the misloaded assembly is modeled with 4.95 wt.% 235U enrichment and no burnup. Eccentric positioning option 3, shown in Figure 6-7 is used for this analysis. Section 6.5.4 demonstrates that this is the most conservative option for a nominal fuel loading. The seven misloaded FA locations ensure that the most conservative misloaded fuel assembly configuration is analyzed.

The fuel compositions are calculated with TRITON/ORIGEN-S including the presence of a neutron source assembly. The results are shown in Table 6-27. The highest value of k95/95 at a boron concentration of 1450 ppm [

Table 6-27 Single Misloaded Fuel Assembly Results Map # Boron (ppm) keff k k95/95 Map 6 1450 [ ] Map 7 1450 [ ] Map 8 1450 [ ] Map 9 1450 [ ] Map 10 1450 [ ] Map 11 1450 [ ] Map 12 1450 [ ] 6.5.11 Multiple Misloaded Fuel Assemblies [

The fuel loading map shown in Figure 6-23 is used for the first configuration. The Zone 1 fuel is modeled with the limiting enrichment and burnup as shown in Table 6-15. A uniform axial burnup shape is used as well as the conservative axial burnup shape at [ ] GWD/MTU as shown in Table 6-6. The nominal loading map, shown in Figure 6-6, is used for the second configuration where the Zone 1 fuel is modeled at [

                                ] burnup. A uniform axial burnup shape is used as well as the conservative axial burnup shape at [          ] GWD/MTU as shown in Table 6-6.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Both configurations are modeled with the eccentric fuel positioning scenarios 1 through 4, as shown in Figure 6-25, Figure 6-26, Figure 6-7, and Figure 6-27. Both multiple misload configurations use fuel composition data from TRITON/ORIGEN-S that assume the presence of a neutron source assembly. The multiple misload accident is the most reactive accident and determines the minimum boron concentration in the spent fuel pool. To facilitate the determination of the minimum boron concentration, calculations are performed at soluble boron amounts of 0, 800, 1200, 1600, and 2000 ppm, as well as the final chosen minimum value of 1450 ppm. Plots of keff versus soluble boron concentration are shown in Figure 6-42 and Figure 6-43. Results of k95/95 at the minimum soluble boron concentration of 1450 ppm are shown in Table 6-28. The table shows that at a boron concentration of 1450 ppm the largest value of k95/95 is [ ] © Copyright 2023 by NuScale Power, LLC 205

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-42 Keff versus Boron Concentration for Multiple Misloaded New Fuel Assemblies [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-43 Keff versus Boron Concentration for Multiple Misloaded Fuel Assemblies [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 6-28 Multiple Misloaded Fuel Assembly Results Boron Configuration1 Concentration keff k k95/95 (ppm) Map=13, Shift=1, shape 1450 [ ] Map=13, Shift=2, shape 1450 [ ] Map=13, Shift=3, shape 1450 [ ] Map=13, Shift=4, shape 1450 [ ] Map=13, Shift=1, flat 1450 [ ] Map=13, Shift=2, flat 1450 [ ] Map=13, Shift=3, flat 1450 [ ] Map=13, Shift=4, flat 1450 [ ] Map=5, Shift=1, shape 1450 [ ] Map=5, Shift=2, shape 1450 [ ] Map=5, Shift=3, shape 1450 [ ] Map=5, Shift=4, shape 1450 [ ] Map=5, Shift=1, flat 1450 [ ] Map=5, Shift=2, flat 1450 [ ] Map=5, Shift=3, flat 1450 [ ] Map=5, Shift=4, flat 1450 [ ] The term Shift refers to the eccentric positioning scenario. 6.5.12 Fuel Storage System Seismic Event The mechanical analysis of the FSRs for a seismic event demonstrates that the FSRs may shift a few inches, [

                                                                                            ] The seismic event is analyzed in three parts - FSR motion by itself, storage cell compression by itself, and both effects simultaneously.

The rack motion is conservatively modeled by placing Zone 2 FAs directly across from each other in adjacent FSRs. This is shown in Figure 6-24. This pattern is broken in one location where one Zone 2 FA is replaced by a Zone 1 FA to prevent placing two Zone 2 assemblies directly adjacent to each other in the same FSR. The loading pattern shown in Figure 6-24 conservatively bounds the seismically induced motion because other locations have Zone 2 fuel directly across from each other, which is not possible by lateral motion of the FSRs. [

                                                                                                          ] The gap is reduced [                                                    ] to eliminate possible geometry conflicts in the criticality model. The length of the absorber plates is reduced by [                                       ] to eliminate geometry conflicts. This is a conservative change as the amount of neutron absorber is [
                                                                                     ] use the nominal fuel loading map shown in Figure 6-6. This is the same loading map used in Section 6.5.5 to determine the burnup and enrichment limits.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The FSR motion cases, with and without [ ] are analyzed with eccentric fuel scenarios 1, 3, and 4, as shown in Figure 6-25, Figure 6-7, and Figure 6-27. [ ] cases are analyzed with eccentric fuel position option 3 providing conservative results as shown in Section 6.5.5. The seismic event is analyzed with the whole pool model. The whole pool model [ ] This is conservative as the FSRs could move to close the gap partially or fully during the seismic event. The analysis is performed with Zone 1 fuel at each of the of the enrichment values shown in Table 6-15 and an assembly burnup at the highest integer value less than or equal to the value shown in Table 6-15. Zone 2 fuel is modeled as fresh fuel with a 4.95 wt.% 235U enrichment. The fuel compositions are calculated with TRITON/ORIGEN-S assuming the presence of a neutron source assembly. A uniform axial burnup distribution is used as the analysis in Section 6.5.5 shows this is conservative with respect to the conservative axial burnup shapes from Table 6-6. The analysis is performed at the minimum soluble boron concentration of 1450 ppm. The results are shown in Table 6-30, Table 6-31, and Table 6-32 and are summarized in Table 6-29 below. [

Table 6-29 Summary of Seismic Event Results [

Table 6-30 Seismically Induced Fuel Storage Rack Motion Results Soluble Assembly (1) Assembly (1) delta keff Case Boron Fuel Map Shift Option Burnup Enrichment keff std dev keff k95/95 (abnormal (ppm) (GWD/MTU) (wt.% 235U) - normal) Normal 1450 [ ] Abnormal 1450 [ ] Normal 1450 [ ] Abnormal 1450 [ ] Normal 1450 [ ] Abnormal 1450 [ ]

     © Copyright 2023 by NuScale Power, LLC Table 6-31 [                                             ]

Assembly Assembly delta keff Soluble Boron Enrichment Case Burnup keff k k95/95 (compressed - (ppm) (wt.% 235U) (GWD/MTU) nominal) Nominal 1450 [ ] Abnormal 1450 [ ] Table 6-32 Results for Simultaneous Fuel Storage Rack Shift [ ] Soluble Assembly Ave delta keff Assembly Case Boron Fuel Map Shift Option Burnup keff std dev keff k95/95 (abnormal - Enrichment (ppm) (GWD/MTU) normal) Normal 1450 [ ] Abnormal 1450 [ ] Normal 1450 [ ] Abnormal 1450 [ ] Normal 1450 [ ] Abnormal 1450 [ ] TR-145417-NP NuScale US460 Fuel Storage Rack Design Topical Report 210 Revision 0

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 6.5.13 Temperature Variation The burnup versus enrichment limit determined in Section 6.5.5 as well as the abnormal conditions (dropped assembly, single misloaded FA, multiple misloaded FAs, and seismic event) are modeled at a SFP temperature of [ ] Other pool temperatures are evaluated with the whole pool model to show that the maximum reactivity occurs at [ ] The analysis is performed with Zone 1 fuel at each of the of the enrichment values shown in Table 6-6 and an assembly burnup at the highest integer value less than or equal to the value shown in Table 6-6. Zone 2 is modeled as fuel enriched to 4.95 wt.% 235U with no burnup. The fuel compositions are calculated with TRITON/ORIGEN-S assuming the presence of a neutron source assembly. A uniform axial burnup distribution is used as the analysis in Section 6.5.5 shows this is conservative with respect to the conservative axial burnup shapes from Table 6-6. The cases are performed both with no soluble boron and soluble boron at the minimum soluble boron concentration of 1450 ppm. Temperature is varied by 5°F from [ ] and by 20°F from [ ] with a final point at 211°F. Results are shown in Figure 6-44 and Figure 6-45. © Copyright 2023 by NuScale Power, LLC 211

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-44 Keff versus Spent Fuel Pool Temperature, 0 ppm Boron [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-45 Keff versus Spent Fuel Pool Temperature, 1450 ppm Boron [

The results are summarized in Table 6-33, which shows the value of keff and k for the nominal temperature of [ ] and the maximum values over the temperature range. Some few cases show a very small increase in keff at [ ] with the increase smaller than the standard deviation of the delta k. These small increases are due to statistical fluctuation in the results and are not considered to be indicative of a true maximum. The small increases are neglected. © Copyright 2023 by NuScale Power, LLC 213

Table 6-33 Spent Fuel Pool Reactivity versus Temperature Analysis Temperature Enrichment Burnup Temperature Delta Std Dev of Boron (ppm) keff k of Max. keff Max. keff k (wt.% 235U) (GWD/MTU) (°F) k-effective Delta k (°F) 0 [ ] 0 [ ] 0 [ ] 0 [ ] 0 [ ] 0 [ ] 0 [ ]

     © Copyright 2023 by NuScale Power, LLC 1450       [                                                                                                                ]

1450 [ ] 1450 [ ] 1450 [ ] 1450 [ ] 1450 [ ] 1450 [ ] TR-145417-NP NuScale US460 Fuel Storage Rack Design Topical Report 214 Revision 0

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 6.5.14 Effect of Cooling Time The TRITON/ORIGEN-S cases that provided fuel composition data assumed no decay time. The effect of decay is evaluated through a special set of TRITON/ORIGEN-S cases and KENO V.a cases of the single FSR location model. The TRITON/ORIGEN-S cases used an enrichment of [

                                 ] with decay times ranging from zero to [                                      ]

The enrichment and burnup combination is taken from the limiting values discussed in Section 6.5.5. Fuel temperature, moderator temperature and density are selected from axial node 23 from Table 6-5. The KENO V.a model of a single FSR location uses the same fuel composition for all axial fuel nodes. A single case is run for each decay time. This hybrid model does not accurately represent an FA but does adequately show the effects of decay time on the nuclide list shown in Table 6-1. The results are shown in Figure 6-46, with details of the first 10 days of decay shown in Figure 6-47. [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-46 Keff versus Decay Time After Shutdown [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure 6-47 Keff versus Decay Time After Shutdown, Details to 10 Days [

6.5.15 Neutron Spectrum The neutron spectrum as quantified by the Energy of Average Lethargy of Fission (EALF) is provided by KENO V.a. The values from all cases are compiled with the minimum and maximum values shown in Table 6-34. Table 6-34 Neutron Spectrum (Energy of Average Lethargy of Fission) Case Type Minimum Maximum Single Storage Cell [ ] Whole Pool [ ] Overall [ ] © Copyright 2023 by NuScale Power, LLC 217

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 6.6 Summary of Criticality Evaluations 6.6.1 Fuel Storage Rack Assembly Map FAs are divided into two categories Zone 1: Irradiated fuel with an enrichment dependent burnup limit discussed in Section 6.6.3, Zone 2: Fresh fuel with enrichment limited to 4.95 wt.% 235U with an allowance of 0.05 wt.% 235U. The acceptable locations for fresh and irradiated fuel are shown in Figure 6-6. The fuel zones are shown as 1 and 2. The following rules define this map Rule 1: Zone 1 fuel may be placed anywhere, even in Zone 2 locations. Rule 2: Zone 2 fuel in the outer rows of an individual FSR that face the pool walls must be separated by at least one storage cell. Rule 3: Zone 2 fuel in the outer rows on an individual FSR that faces another FSR:

                  -   Must be separated by at least two storage cells.
                  -   Cannot be directly across from a Zone 2 FA in an adjacent FSR.
                  -   A knights move (two spaces in one direction and one space in the perpendicular direction) is allowed at the corner of an FSR.

Rule 4: At the interface between Rule 2 and Rule 3, Zone 2 fuel may be diagonally adjacent. 6.6.2 Definition of Assembly Enrichment Assembly enrichment is taken as the planar averaged enrichment of the non-gadolinia bearing fuel pins at the axial region where the enrichment is highest. 6.6.3 Burnup Limit versus Assembly Enrichment An assembly is considered as Zone 1 if the assembly average burnup meets the following criteria: 2 BU A x E + B x E + C Equation 6-3 where: Burnup is in units of GWD/MTU E is enrichment in wt.% 235U with an allowance of 0.05 wt.% 235U and [ ] [ ] © Copyright 2023 by NuScale Power, LLC 218

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 [ ] [ ] The burnup limits for selected values of enrichment are shown in Table 6-15, reproduced here for convenience. Table 6-35 Final Burnup versus Enrichment Limit for Zone 1 Enrichment (wt.% 235U) Burnup Limit (GWD/MTU) [ ] [ ] [ ] [ ] [ ] [ ] [ ] 6.6.4 Abnormal and Accident Conditions 6.6.4.1 Multiple Misloaded Fuel Assemblies Two scenarios of multiple misloaded FAs are analyzed as discussed in Section 6.2. This is the most limiting event and sets the minimum soluble boron concentration at 1450 ppm. For this condition the largest value of k95/95 is [ ] well below the limit of 0.95. 6.6.4.2 Single Misloaded Fuel Assembly The single misloaded FA is discussed in Section 6.5.10. The FSRs are fully loaded according to the storage map discussed in Section 6.6.3. The misloaded FA is placed in several locations to ensure the most reactive position is found. Credit is taken for the minimum boron concentration of 1450 ppm. The highest value of k95/95 provides significant margin to the limit of 0.95. 6.6.4.3 Fuel Assembly Dropped On the Fuel Storage Racks An FA dropped on top of the FSRs is separated from the fuel region of the FAs stored in the rack by more than [ ] This distance is sufficient that the dropped assembly is considered decoupled from the fuel stored in the rack. No further consideration is required. 6.6.4.4 Fuel Assembly Mislocated Beside the Fuel Storage Racks An FA dropped beside the array of FSRs is discussed in Section 6.5.9. The FSRs are fully loaded according to the storage map discussed in Section 6.6.3. Several locations are analyzed to ensure the maximum coupling between the mislocated © Copyright 2023 by NuScale Power, LLC 219

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 assembly and the FSR is found. Credit is taken for the minimum soluble boron concentration of 1450 ppm. The maximum value of k95/95 provides a significant margin to the limit of 0.95. 6.6.4.5 Seismic Event A seismic event may move the FSRs [

                                                                              ] These phenomena are analyzed individually, as if they happen at different points in time, and together, as if they occur simultaneously. While the actual FSR motion is only a few inches with the FSRs moving mostly in unison, the analysis assumes the worst possible alignment of FSRs in both horizontal directions at the same time. [
                             ] Credit is taken for the minimum soluble boron concentration of 1450 ppm. The maximum value of k95/95 provides significant margin to the limit of 0.95.

6.6.4.6 Spent Fuel Pool Temperature The criticality analysis is performed at [ ] with the water density slightly above [ ] to account for the submerged depth of the FSRs. Except for very small statistical fluctuations in the KENO V.a results, the FSRs are most reactive at [ ] 6.6.5 Design Parameters 6.6.5.1 Neutron Source Assembly The neutron source assembly is conservatively modeled as a void in the full length of the guide and instrument tubes. The KENO V.a cases that determine the burnup versus enrichment limit and that analyze the various abnormal and accident conditions assume that the irradiated fuel contains an NSA while resident in the core. The same KENO V.a cases also conservatively assume that the NSA is not present in the fuel while in the FSRs. Thus, there is no limitation on the use of neutron source assemblies. 6.6.5.2 Damaged Fuel Assemblies The system bias term that accounts for cladding damage (Section 6.5.3.3) assumes that all fuel rods in all the fresh fuel assemblies are flooded with water. Thus, a fresh FA loaded into a Zone 2 location may have damaged cladding with no additional restriction on number or location of damaged assemblies. © Copyright 2023 by NuScale Power, LLC 220

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 6.6.5.3 Fuel and Burnable Poison Design Fuel assemblies placed in the fuel FSRs must comply with the fuel description provided in Table 6-2. Additionally, only gadolinia is allowed as a burnable absorber. 6.6.5.4 Boron-10 Loading in Neutron Absorber Plates The criticality analysis assumes a Boron-10 loading of [ ] in the neutron absorber plates. The absorber material specified for construction must meet this minimum value while considering manufacturing variations. © Copyright 2023 by NuScale Power, LLC 221

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 7.0 Materials Analysis The designated structural and neutron absorber materials are evaluated for suitability in the established environments and their ability to meet applicable design requirements. Structural materials selected for the FSR are [

                                                                                                     ] Weld filler metals used in the FSR are corrosion resistant [                        ] that are compatible with the base metals in accordance with Part C, Specifications for Welding Rods, Electrodes and Filler Metals to Section II, Materials of the 2017 Edition of the ASME B&PV Code (Reference 10.2.a.ii). There are no bolts in the FSR.

The neutron absorber material selected is MAXUS, a product of Nikkeikin ACT, a wholly owned subsidiary of Nippon Light Metal Holdings. Comparison of the US460 SFP chemistry and the suggested chemistry parameters for a typical Pressurized Water Reactor (PWR) SFP shown in Table 7-1 indicates that the expected chemistry conditions in the US460 SFP are like those in a typical PWR SFP. Table 7-1 NuScale vs Typical Pressurized Water Reactor Spent Fuel Pool Water Chemistry Requirements NuScale SFP Typical PWR SFP Criterion Chemistry Limits Chemistry Limits1 pH [ ] Variable Boron [ ] Variable Conductivity [ ] Variable Chloride Ions [ ] < 0.15 ppm Fluoride Ions [ ] < 0.15 ppm Sulfate Ions [ ] < 0.15 ppm 1 Parameters taken from the EPRI report, Pressurized Water Reactor Primary Water Chemistry Guidelines, Volume 1, Revision 7 (Reference 10.30). 7.1 Material Requirements Requirements on the materials selected for the FSR are grouped into two categories: those for the metallic, or structural, materials and those for the neutron absorber material. Requirements for the structural materials used in the US460 FSR are: Material specifications shall be found in Part A, Ferrous Material Specifications of Section II, Materials, to the ASME B&PV Code, 2017 Edition (Reference 10.2.a.i). Materials shall be capable of an 80-year design life in the US460 SFP environment. Materials shall have been successfully deployed in LWR environments. © Copyright 2023 by NuScale Power, LLC 222

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Materials shall be compatible with the US460 SFP thermal, chemistry and radiation environment. Requirements for the neutron absorber material used in the US460 FSR are: Material shall be qualified for use in an SFP. Potential failure mechanisms that could cause dispersal of the neutron absorber shall be identified and dispositioned. The ability of the neutron absorber to provide the amount of absorption assumed in the criticality analysis of the SFP shall be established. 7.2 Structural Material Evaluation The structural materials for the FSR, [

                                                              ] are provided in Section II, Part A of the ASME B&PV Code (Reference 10.2.a.i). [
                                                                               ]

These materials are selected to be compatible with the new and spent fuel and with the SFP water environment for a minimum of 80 years. Explicit test data for these materials necessary to extrapolate to an 80-year lifetime is not available but these materials have been extensively used in operating LWR reactor internals and SFPs and have demonstrated acceptable performance. [ ] are used extensively in FAs, plant piping, reactor pressure vessel internals, and SFPs that are subjected to environments like, and in some cases more aggressive than, the US460 SFP without significant degradation due to general or galvanic corrosion mechanisms. The most significant risk of degradation to the structural materials in the FSR is corrosion. The materials selected are strongly resistant to corrosion in the environmental conditions as stated above. [ ] demonstrate exceptional corrosion resistance in typical PWR environments. Since the US460 water chemistry aligns well with that used in PWR SFPs that follow the EPRI chemistry guidelines (Reference 10.30) per Table 7-1, it is expected that these corrosion mechanisms do not pose a failure/degradation risk to the materials chosen, provided the EPRI primary water chemistry guidelines for PWR SFPs are followed. Preserving these conditions reduces the likelihood of general corrosion. There is no risk of galvanic corrosion between the [ ] components or with the materials used in a fuel assembly (zirconium alloys, stainless steel alloys and nickel base alloys): these materials are regularly in contact without significant observed galvanic corrosion. Irradiation Embrittlement (IE) is possible for [

                             ] in PWR RV internals screen in for IE susceptibility at a dose of 1E21 n/cm2 (E > 1.0 MeV) ( 1.5 dpa) and [                                      ] welds screen in 2

at a dose of 6.7E20 n/cm (E > 1.0 MeV) ( 1 dpa) according to the EPRI Materials Reliability Program (Reference 10.31). Criticality safety analysis of the FSR array in the US460 SFP shows that the fuel in the FSRs remains subcritical (k95/95 < 0.95 with a © Copyright 2023 by NuScale Power, LLC 223

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 minimum boron concentration of 1450 ppm) under abnormal and accident conditions. Therefore, the driver of neutron flux in the SFP is expected to be the radioactive decay of the fuel. Additionally, a review of two PWR subsequent license renewal applications indicates the following maximum projected fluence values in the reactor pressure vessel: 1.08E20 n/cm2 (E > 1.0 MeV) for 80 years (72 Effective Full Power Years) (Turkey Point Nuclear Plant Units 3 and 4 Subsequent License Renewal Application (Reference 10.32)) 7.26E19 n/cm2 (E > 1.0 MeV) for 80 years (68 Effective Full Power Years) at the base metal clad interface (Surry Power Station Units 1 and 2 Application for Subsequent License Renewal (Reference 10.33)) By engineering judgment based on a comparison of the conditions in the US460 SFP and in the reactor pressure vessel at full power, (e.g., temperature, driver of neutron flux, critical/subcritical conditions), and given that multiple PWRs have a maximum reactor pressure vessel 80-year fluence significantly below the screening criterion, IE susceptibility is not expected. Temperature dependence of Stress Corrosion Cracking (SCC) in sensitized [

                            ] is tested in oxygenated PWR primary water as documented in the National Association of Corrosion Engineers paper, The Effects of Boric Acid, Solution Temperature, and Sensitization on SCC Behavior Under Elevated Temperature Water (Reference 10.34), and no cracking is observed below 200ºF. This is consistent with historical research documented in EPRI report, Non-Class 1 Mechanical Implementation Guideline and Mechanical Tools (Reference 10.35), which found SCC is observed above 200ºF in high purity (i.e., low sulfates and halogens; less than 0.15 ppm) oxygenated water (oxygen > 0.1 ppm).

Transgranular SCC can occur at temperatures from approximately 125ºF to approximately 200ºF for sensitized [ ] and dissolved oxygen concentrations of approximately 0.006 ppm to approximately 0.07 ppm, respectively. However, weld sensitization of the fuel rack components is not expected to occur due to the use of [ ] Though oxygen concentration is not limited by the US460 SFP chemistry requirements, due to the combination of the SFP operating temperature and use of low carbon [

                            ], sensitized material and SCC are not expected.

[ ] is susceptible to SCC in a primary water environment due to stresses from cold work or due to Weld Residual Stress (WRS). The [

                     ] structural materials used in the fabrication of the US460 FSRs are supplied in an annealed condition and are not cold worked. Therefore, SCC due to cold work is not expected. [                                            ] specification is introduced to reduce general SCC susceptibility resulting from material sensitization during welding. While some level of WRS is present as a result of fuel rack construction, due to the combination of controlled SFP water chemistry and the use of a less susceptible material, SCC due to WRS is not expected. Chloride-induced SCC is not expected to be a concern given the tight chloride controls placed on the US460 SFP as shown in Table 7-1. [
                            ] also exhibits a reduced susceptibility to general corrosion. If the

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 proposed water chemistry is maintained and no contamination of the rack materials occurs, SCC is not expected to occur. The only sustained stress the FSRs are subject to is deadweight. The maximum stress in the FSR due to deadweight is approximately [ ] which is well below the stress level that SCC typically initiates in components made of [ ] material in a controlled water chemistry. The FSR support legs are [

                                             ] fabricated from [                                        ]

material. [ ] they are assessed for the potential impact of boric acid corrosion. A study of American Iron and Steel Institute [ ] under the following conditions is documented in EPRIs Boric Acid Corrosion Guidebook (Reference 10.36): Test periods of two or four weeks Testing at 70ºF with boron concentration of 50,000 ppm Testing at 140ºF with boron concentration of 130,000 ppm This study found essentially no corrosion of the [ ] in any of the tests. As the test conditions have a boron concentration of more than an order of magnitude greater than the US460 SFP water chemistry shown in Table 7-1, boric acid corrosion is not expected to occur. Other possible degradation sources, such as contamination events or manufacturing defects, are outside the scope of this evaluation. 7.3 Neutron Absorber Material Evaluation The neutron absorber used in the US460 FSR is MAXUS, a high-density, clad neutron absorber material consisting of a neutron absorbing core with 20 to 40 percent weight of natural grade boron carbide (B4C) in an AA 1070 aluminum matrix, surrounded by AA 5000 series aluminum cladding. MAXUS is described in EPRIs Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Storage and Transportation Applications (Reference 10.37). The neutron absorbing core is a high purity (99.7%) aluminum matrix with a uniform distribution of B4C particles in the matrix leading to a near theoretical density. Accelerated long-term corrosion testing of MAXUS flat and bent coupons has been performed in simulated boiling water reactor and PWR spent fuel pool environments for 18,000 hours at 195ºF, and for an additional 40,000 hours as described in EPRIs, Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Storage and Transportation Applications (Reference 10.37). The corrosion rate during these exposures is determined to be essentially negligible, and the 10B areal density did not statistically change during the qualification testing. Limited pitting is observed in some flat coupons during the long-term testing after five years, but otherwise the flat and bent © Copyright 2023 by NuScale Power, LLC 225

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 coupons exhibit no substantial corrosion. Areal density during the long-term testing is not affected for either coupon type. Aside from corrosion, MAXUS in the US460 SFP is subject to irradiation. Due to the widespread use of Al-B4C neutron absorbers throughout the nuclear industry and the associated qualification testing performed on those materials, it is well known this family of material is not substantially affected by radiation exposure. Since the relevant properties of MAXUS are sufficiently like those of these other materials, it is concluded that MAXUS exhibits similar acceptable performance relative to its irradiation behavior. MAXUS has previously been approved by the NRC for use in a PWR SFP as discussed in the NRCs Safety Evaluation of the Arizona Public Service Companys request to use MAXUS inserts in the Palo Verde Nuclear Operating Station FSRs (Reference 10.38). 7.4 Material Evaluation Conclusions The [ ] material selected for the US460 FSR has a proven positive performance in LWRs and is expected to perform well in the US460 SFP environment for the design lifetime of 80 years. Accelerated long-term testing of the neutron absorber material, MAXUS, resulted in essentially negligible corrosion rates and no effect on 10B areal density. MAXUS is sufficiently like other Al-B4C neutron absorbers in relevant properties to conclude that MAXUS is not affected by irradiation. Therefore, the materials selected for the FSR are suitable for use in the SFP environment. © Copyright 2023 by NuScale Power, LLC 226

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 8.0 Manufacturing, Operation, and Maintenance 8.1 Manufacturing Fabrication of the FSRs is controlled using traveler documentation following industry standard construction procedures. Fabrication shall be performed in accordance with the fabrication vendors Quality Program that shall be reviewed and approved by the FSR designer. Material shall be procured from vendors on the FSR designers approved suppliers list, and construction shall be monitored with a series of hold points, with steps requiring Non-Destructive Examination being subject to the guidance of NF-5300 in Subsection NF to Section III, Division 1, of the ASME B&PV (Reference 10.2.b.i). Each FSR shall meet Class B cleanliness requirements per 302.2, Subpart 2.1, of ASME NQA-1-2008 (Reference 10.5), and shall be packaged, shipped, and stored per the Level C requirements set forth in Subpart 2.2 of ASME NQA-1-2008 (Reference 10.5). 8.2 Operations and Maintenance of the Fuel Storage Rack The material selection, structural, thermal-hydraulic, and criticality safety analyses described in previous sections of this report demonstrate that the structural design of the FSR is suitable for the SFP environment, including consideration of load drop and seismic events. Based on these results, maintenance of the FSR structural components after they have been put in service is not required. Monitoring of the fixed neutron absorbers contained in the FSR is performed to the requirements of NEI 16-03-A, Revision 0, Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools (Reference 10.39). The neutron absorber monitoring program can identify whether changes to the material are occurring, and the anticipated characteristics of change are verified. 8.2.1 Neutron Absorber Monitoring Requirements The neutron absorber monitoring program consists of a neutron absorber coupon tree with periodic removal and testing of neutron absorber coupons and SFP water chemistry monitoring. The coupon testing program shall consist of the following elements. The number of coupons is sufficient to provide sampling at an appropriate interval for the intended life of the neutron absorber. Sampling intervals for new materials that do not have applicable operating experience in conditions like the US460 pool environment is 5 years initially, with subsequent intervals up to 10 years being acceptable. For well-known materials that have been used for several years in conditions like the US460 pool environment, initial and subsequent intervals up to 10 years are acceptable. Coupon testing is categorized as a combination of basic and full testing used to identify whether unanticipated changes are occurring. The extent to which each of © Copyright 2023 by NuScale Power, LLC 227

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 these is utilized is determined based upon the Operating Experience (OE) of the material.

                  -    Basic testing consists of visual observations, dimensional measurements, and weight that is performed at the SFP and is appropriate when previous testing and OE of the material indicates that there are no degradation mechanisms that would result in loss of 10B areal density that would affect reactivity. Basic testing is to occur at least every ten years.
                  -    Full testing may consist of a combination of mass-density measurements, 10B areal density measurements, microscopic analysis, and characterization of changes, in addition to the basic testing parameters. The 10B areal density measurement is to occur at least every ten years.

The locations of the coupons are such that their exposure to parameters controlling change mechanisms (e.g., gamma fluence, temperature) is conservative or equivalent to the in-service neutron absorbers. Results are acceptable to confirm the continued performance of neutron absorber materials if:

                  -    For materials that are not anticipated to have a loss of 10B areal density; the 10B areal density of the test coupon is the same as its original 10B areal density (within the uncertainty of the measurement).
                  -    For materials that are anticipated to have a loss of 10B areal density; the 10B areal density of the test coupon is greater than the 10B areal density used in the criticality analysis.

8.2.2 Evaluating Neutron Absorber Test Results For coupon testing, results from neutron absorber monitoring fall within the broad categories of: confirmation that no material changes are occurring confirmation that anticipated changes are occurring identification that unanticipated changes are occurring Relevant processes are used to evaluate results of the monitoring program with the criticality analysis input. If there are no changes, or if anticipated changes are occurring that have already been accounted for, then the material condition continues to be adequately represented in the criticality analysis. If unanticipated changes are identified (either new mechanisms or anticipated mechanisms at rates or levels beyond those anticipated), then additional actions are necessary. In addition to relevant regulatory and licensing processes (e.g., corrective action program, reporting requirements, the 10 CFR 50.59 process, operability determination or functionality assessment), the following technical evaluations are necessary. Determine if unanticipated changes could result in a loss of 10B areal density. © Copyright 2023 by NuScale Power, LLC 228

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Determine if unanticipated changes not resulting in loss of 10B areal density have an impact on criticality analyses. 8.2.3 Neutron Absorber Coupons Neutron absorber coupons shall be taken from the same production lot(s) as the neutron absorber plates installed in the FSRs. Prior to placement in the SFP, the coupons shall be pre-characterized to create a baseline. Coupons shall be permanently and uniquely marked to maintain traceability to production lots of MAXUS used in the fuel racks. The permanent markings shall ensure a Quality Assurance program level of record for the coupons. Prior to immersion in the SFP, each coupons weight, dimensions, and 10B areal density shall be measured and recorded, and each coupon shall be photographed. Thirty-six coupons shall be immersed in the SFP, and four coupons shall be archived (i.e., not immersed in the SFP) (Table 8-1). The coupon types included in the monitoring program are general, which are coupons in the manufactured state (i.e., the same condition as the MAXUS installed in the fuel racks), and coupons in a galvanic couple with [ ] which is used in construction of the fuel racks. Table 8-1 Neutron Absorber Coupon Type and Count SFP Coupon Type Archive Coupons Total Coupons Coupons General 24 2 26 MAXUS-[ 12 2 14

                  ] (Galvanic Couple)

The neutron absorber coupons shall be placed in a coupon tree within the SFP. The location of the coupons in the SFP shall be such that their exposure to parameters controlling change mechanisms (e.g., gamma fluence, temperature) is conservative or equivalent to the in-service neutron absorbers. A portion of the in-service fuel rack neutron absorber panels shall be visually examined in-situ by camera at every refueling outage for the first module in service for 10 years following commissioning of the FSRs, then approximately every five years after 10 years with acceptable performance (Table 8-2). Additionally, coupons in the coupon tree shall be visually examined in-situ at each refueling outage for the first module in service to monitor for bubbling, blistering, corrosion pitting, cracking, discoloration, or flaking. © Copyright 2023 by NuScale Power, LLC 229

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table 8-2 Neutron Absorber Material in Spent Fuel Pool - Inspection Type and Schedule No. of In-Situ Visual Interval MAXUS Component Full Testing Schedule (Years) Coupons (Years) General Coupon Every refueling outage for 2 coupons after 5 and 10 years 24 the first module in in service, then 2 coupons service(1) every 10 years thereafter MAXUS- [ Every refueling outage for 1 couple after 5 and 10 years in

                                  ]      12     the first module in          service, then 1 couple every service(1)                   10 years thereafter Fuel Rack Neutron Absorber                       Every refueling outage for Panels                                           the first module in service(1) for 10 years, then approximately every 5 years after 10 years of acceptable performance.

N/A N/A Without acceptable performance, the next visual inspection shall occur at an interval established by engineering evaluation. 1 Since the US460 plant design contains multiple NuScale Power Modules each with its own refueling interval that share the same FSR, visual examination is linked to the first module that is placed in service. At approximately 5, 10, 20, 30, 40, 50, 60, 70 and 80 years from commissioning of the spent FSRs, two general coupons and one galvanic couple coupon shall be removed (Table 8-2). The removed coupon(s) shall be visually examined, weighed, measured, and 10B areal density shall be measured, with the following acceptance criteria: 10B areal density of 95% of pre-immersion measurement Thickness increases due to general corrosion 20% compared to pre-characterization measurement Pitting corrosion not exceeding a rating of A-2, B-2, C-2 per ASTM G46 (Reference 10.40) paragraph 7.2.1 is acceptable; pitting that exceeds this rating is acceptable if that area of greatest pitting satisfies the criteria below:

                  -    Edge corrosion shall not reduce the coupon length or width by more than 1/16 inch If any of the acceptance criteria are not satisfied, an engineering evaluation shall be performed to determine if there is any effect on the performance of the FSR functions, and if any corrective action is required. Coupons that are not destroyed are suitable for return to the pool for continued use in the monitoring program. Coupons that are returned to the pool for continued monitoring either supplement new coupons for future inspections or serve as a primary coupon. This determination is based on evaluation of coupon monitoring results history.

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The testing schedule described in this section requires a total of 20 general coupons and 10 galvanic couple coupons for an 80-year service life. Four additional general coupons and two additional galvanic coupons are included in Table 8-2 for testing after 80 service years to permit off-loading the spent fuel during decommissioning. 8.2.4 Neutron Absorber Coupon Tree The neutron absorber coupon tree is designed to securely hold the coupons identified in Table 8-2 in the SFP as part of the neutron absorber monitoring plan. Requirements for the coupon tree are as follows. Placed in the SFP such that the coupons exposure to parameters controlling change mechanisms (e.g., gamma fluence, temperature) is conservative or equivalent to the in-service neutron absorbers in the SFP racks. Service location allows visual inspection of in-situ general coupons. Movement or removal within the SFP is achieved using a specialized tool; movement cannot be achieved with hooks, grapples, or other general tools. Uniquely identified as equipment vital to U.S. Nuclear Regulatory Commission (NRC) monitoring requirements. Coupon Tree or Do Not Disturb should be engraved in the upper surface of the coupon tree. These markings should be visible from the fuel handling bridge. The coupons are secured while the coupon tree is being inserted or removed from the SFP and during the coupon tree service time in the SFP. 8.2.5 Spent Fuel Pool Water Chemistry Monitoring Water chemistry target limits for the US460 SFP are compared to EPRI PWR primary water chemistry guidelines in Section 7.0 and it is shown these chemistry conditions are similar. The following pertinent water chemistry parameters are recommended to be monitored at the noted frequencies: pH and Conductivity: once a week by pH/conductivity meter Boron: once a day or per Tech Specs when SFP is credited for Shutdown Margin (SDM) OR once a week if SFP is not credited for SDM by titration Chloride, Fluoride, and Sulfate: once a month by Ion Chromatography (IC) 8.2.6 Operations and Maintenance Summary An operations and monitoring plan for the fixed neutron absorbers in the FSR is developed to be consistent with the guidance provided in NEI 16-03-A, Revision 0 (Reference 10.39). This plan includes pre-service and in-service testing plans for the neutron absorbing material used in the US460 FSRs. A total of twenty-six general coupons and fourteen galvanic couple coupons comprises the testing program. Twenty-four general coupons and twelve galvanic coupons are placed in a coupon tree in the SFP while two general coupons and two galvanic coupons are stored as archive coupons. © Copyright 2023 by NuScale Power, LLC 231

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Coupons shall be subject to pre-characterization measurements for weight, dimensions, and 10B areal density. Additionally, coupons shall be photographed as part of the pre-characterization process. Two general coupons and one galvanic coupon shall be removed from the SFP at approximately 5, 10, 20, 30, 40, 50, 60, 70, and 80 years from commissioning of the spent FSRs. These coupons shall be visually examined, weighed, measured, and 10B areal density shall be measured, with the following acceptance criteria: 10B areal density of 95% of pre-immersion measurement Thickness increases due to general corrosion 20% compared to pre-characterization measurement Pitting corrosion not exceeding a rating of A-2, B-2, C-2 per ASTM G46 (Reference 10.40) paragraph 7.2.1 is acceptable; pitting that exceeds this rating is acceptable if that area of greatest pitting satisfies the criteria below:

                  -    Edge corrosion shall not reduce the coupon length or width by more than 1/16 inch A portion of the in-service fuel rack neutron absorber panels shall be visually examined in-situ by camera at every refueling outage of the first module in service for 10 years following commissioning of the FSRs, then approximately every five years after 10 years of acceptable performance. Without acceptable performance, the next visual inspection shall occur at an interval established by engineering evaluation. The visual examination is to be performed to monitor for bubbling, blistering, corrosion pitting, cracking, discoloration, or flaking. Additionally, coupons in the coupon tree shall be visually examined in-situ at each refueling outage of the first module in service to monitor for bubbling, blistering, corrosion pitting, cracking, discoloration, or flaking.

NuScale has established water chemistry target limits which are like EPRI primary water chemistry requirements. The following pertinent water chemistry parameters are recommended to be monitored at the noted frequencies: pH and Conductivity: once a week by pH/conductivity meter, Boron: once a day or per Tech Specs when SFP is credited for SDM OR once a week if SFP is not credited for SDM by titration, Chloride, Fluoride, and Sulfate: once a month by IC. © Copyright 2023 by NuScale Power, LLC 232

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 9.0 Summary and Conclusions This report presents design and analysis of the US460 FSR for storing unirradiated and irradiated NuFuel-HTP2' fuel. This report examines the structural, thermal hydraulic, criticality and material performance of the FSR design. Structural analyses examine the mechanical integrity of the FSR design considering deadweight, impacts due to a dropped fuel assembly, and loading from an SSE in accordance with the Design Specific Review Standard, Section 3.8.4 (Reference 10.1.b). Structural analyses additionally examine stresses due to lifting and the forces exerted in the event of a fuel assembly stuck in the rack. Thermal hydraulic analyses demonstrate that flow through the FSR is adequate for decay heat removal during anticipated operating occurrences and accident conditions and that natural circulation is sufficient to prevent nucleate boiling during anticipated operating conditions. Criticality analyses ensure that criticality control is maintained for normal and credible abnormal conditions in accordance with Standard Review Plan, Section 9.1.1 (Reference 10.41). The analysis defines acceptable storage configurations for unirradiated and irradiated fuel that ensures criticality control is maintained. Evaluation of the structural and neutron absorbing materials assesses the chemical compatibility of the materials with the spent fuel pool environment. The analyses and evaluations demonstrate that the FSR design complies with applicable requirements and that the FSR design is acceptable for use in operating plants that utilize the US460 Standard Plant Design. This report defines acceptable storage configurations for unirradiated and irradiated nuclear fuel that ensures criticality control is maintained for normal and credible conditions in the spent fuel pool. This report also describes the quality programs applicable to manufacturing of the FSR and provides information concerning operation and monitoring of the FSRs. The operation and monitoring information provides the basis for development of FSR monitoring programs for licensees that adopt this report. © Copyright 2023 by NuScale Power, LLC 233

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 10.0 References 10.1 United States Nuclear Regulatory Commission Design-Specific Review Standard Review for NuScale SMR Design, United States Nuclear Regulatory Commission, Rockville, MD:

a. Section 3.7.1, Revision 0, Seismic Design Parameters, June 2016. ADAMS Accession No. ML15355A384.

b. Section 3.8.4, Revision 0, Other Seismic Category I Structures, June 2016.

ADAMS Accession No. ML15355A444.

c. Section 9.1.2, Revision 0, New and Spent Fuel Storage, June 2016. ADAMS Accession No. ML15356A584.

10.2 American Society of Mechanical Engineers Boiler and Pressure Vessel Code, American Society of Mechanical Engineers, New York, NY, 2017 Edition (July 2017):

a. Section II, Materials i) Part A, Ferrous Material Specifications ii) Part C, Specifications for Welding Rods, Electrodes and Filler Metals iii) Part D, Properties (Customary)

b. Section III, Rules for Construction of Nuclear Facility Components, Division 1 i) Subsection NF, Supports ii) Nonmandatory Appendix F, Rules for Evaluation of Service Loadings with Level D Service Limits

c. Section V, Nondestructive Examination,

d. Section IX, Welding and Brazing Qualifications.

10.3 Title 10 of the Code of Federal Regulations, Part 50, Domestic Licensing of Production and Utilization Facilities, Office of the Federal Register, National Archives and Records Administration, Washington, D.C.:

a. Section 50.68, Criticality Accident Requirements

b. Appendix A, General Design Criteria for Nuclear Power Plants, last updated August 28, 2007 i) General Design Criterion 1, Quality Standards and Records ii) General Design Criterion 2, Design Bases for Protection Against Natural Phenomena iii) General Design Criterion 61, Fuel Storage and Handling and Radioactivity Control iv) General Design Criterion 62, Prevention of Criticality in Fuel Storage and Handling

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

c. Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, last updated November 18, 2019.

10.4 United States Nuclear Regulatory Commission Regulatory Guide 1.29, Revision 6, Seismic Design Classification for Nuclear Power Plants, United States Nuclear Regulatory Commission, Rockville, MD, July 2021. ADAMS Accession No. ML21155A003. 10.5 American Society of Mechanical Engineers NQA-1-2008, Quality Assurance Requirements for Nuclear Facility Applications, including NQA-1a-2009 (2008 Edition with 2009 Addenda), American Society of Mechanical Engineers, New York, NY, August 2009. 10.6 ANSYS Mechanical, Finite Element Analysis Software for Structural Analysis, Version 2019 R2 (19.2), September 2018, ANSYS Inc, Canonsburg, PA. 10.7 ANSYS LS-DYNA, Multiphysics Solver, Version R10.1, February 2018, ANSYS Inc, Canonsburg, PA. 10.8 SCALE, Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, Version 6.2.4, June 2020, Oak Ridge National Laboratory, Oak Ridge, TN. 10.9 Letter from NuScale Power, LLC to U.S. Nuclear Regulatory Commission, "Submittal of Technical Reports Supporting the NuScale Design Certification Application (NRC Project No. 0769), dated December 30, 2016, United States Nuclear Regulatory Commission, Rockville, MD. ADAMS Accession No. ML17005A112. [NuScale Report TR-0816-49833-P, Revision 0, Fuel Storage Rack Analysis, issued September 2016, was submitted by this letter]. 10.10 Nuclear Energy Institute document NEI 12-16, Revision 4, Guidance for Performing Criticality Analyses of Fuel Storage at Light Water Reactor Power Plants, Nuclear Energy Institute, Washington, D.C., September 2019. ADAMS Accession No. ML19269E069. 10.11 United States Nuclear Regulatory Commission Regulatory Guide 1.240, Revision 0, Fresh and Spent Fuel Pool Criticality Analyses, United States Nuclear Regulatory Commission, Rockville, MD, March 2021. ADAMS Accession No. ML20356A127. 10.12 American Society of Materials International, Atlas of Stress-Strain Curves, American Society of Materials International , Russell Township, OH, 2nd Edition (2002). 10.13 K.C. Radil and A.B. Palazzolo, National Aeronautics and Space Administration Technical Memorandum 106485, Influence of Temperature and Impact Velocity on the Coefficient of Restitution, NASA Lewis Research Center, Cleveland, OH, July 1994. © Copyright 2023 by NuScale Power, LLC 235

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 10.14 United States Nuclear Regulatory Commission Regulatory Guide 1.61, Revision 1, Damping Values for Seismic Design of Nuclear Power Plants, Rockville, MD, March 2007. ADAMS Accession No. ML070260029. 10.15 American National Standards Institute Standard N14.6, Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4,500 kg) or More, American National Standards Institute, Washington DC, January 1993. 10.16 I.E. Idelchik, Handbook of Hydraulic Resistance, Begell House Inc, Danbury, CT, 3rd Edition (1994). 10.17 ANSYS Fluent, Fluid Simulation Software, Version 2021/R2, July 2021, ANSYS Inc, Canonsburg, PA:

a. Theory Guide

b. Users Manual.

10.18 Warren M. Rohsenow, James P. Hartnett, and Young I. Cho, Handbook of Heat Transfer, McGraw-Hill Education, New York, NY, Third Edition (1998). 10.19 United States Nuclear Regulatory Commission NUREG/CR-6665, Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel, United States Nuclear Regulatory Commission, Rockville, MD, February 2000. ADAMS Accession No. ML003688150. 10.20 United States Regulatory Commission Regulatory Guide 1.183, Revision 0, Alternate Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, United States Nuclear Regulatory Commission, Rockville, MD, July 2000. ADAMS Accession No. ML003716792. 10.21 Pacific Northwest National Laboratory PNNL-18212, Revision 1, Update of Gas Release Fractions for Non-LOCA Events Utilizing the Revised ANS 5.4 Standard, Pacific Northwest National Laboratory, Richland, WA, June 2011. 10.22 Framatome Topical Report BAW-102331(NP)-A, Revision 1, COPERNIC Fuel Rod Design Code, Framatome ANP, Lynchburg, VA, January 2004. ADAMS Accession No. ML042930240. 10.23 United States Nuclear Regulatory Commission NUREG/CR-6698, Guide for Validation of Nuclear Criticality Safety Calculation Methodology, United States Nuclear Regulatory Commission, Rockville, MD, December 2000. 10.24 Owen, D.B., Sandia Corporation Monograph SCR-607, Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation, Albuquerque, NM, March 1963. © Copyright 2023 by NuScale Power, LLC 236

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 10.25 Royston, Patrick, Approximating the Shapiro-Wilk W-test for non-normality, Statistics and Computing, 1992. 10.26 American National Standards Institute ANSI N15.15-1974, NSI N15.15-1974, Assessment of the Assumption of Normality (Employing Individual Observed Values), American National Standards Institute, Washington DC, October 1973 (Reaffirmed 1981). 10.27 Nuclear Energy Agency, International Handbook of Evaluated Criticality Safety Benchmark Experiments, Nuclear Energy Agency, Paris, France, September 2021. 10.28 United States Nuclear Regulatory Commission NUREG/CR-6361, Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages, United States Nuclear Regulatory Commission, March 1997. 10.29 United States Nuclear Regulatory Commission NUREG/CR-7109, An Approach for Validating Actinide and Fission Product Burnup Credit Criticality Safety Analyses-Criticality (keff) Predictions (NUREG/CR-7109, ORNL/TM-2011/514), United States Nuclear Regulatory Commission, Rockville, MD, April 2012. 10.30 Pressurized Water Reactor Primary Water Chemistry Guidelines, Volume 1, Revision 7, Electric Power Research Institute, Palo Alto, CA. 2014. Report 3002000505. 10.31 Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175, Revision 1), Electric Power Research Institute, Palo Alto, CA. 2017. Report 3002010268. 10.32 United States Nuclear Regulatory Commission Document, Turkey Point Nuclear Plants Units 3 and 4 Subsequent License Renewal Application, Revision 1 (Public Version), United States Nuclear Regulatory Commission, Rockville, MD, April 2018. ADAMS Accession No. ML18113A146. 10.33 United States Nuclear Regulatory Commission Document, Surry Power Station Units 1 and 2 Application for Subsequent License Renewal, United States Nuclear Regulatory Commission, Rockville, MD, October 2018. ADAMS Accession No. ML18291A828. 10.34 K. Arioka et al., The Effects of Boric Acid, Solution Temperature, and Sensitization on SCC Behavior Under Elevated Temperature Water, Corrosion 83, NACE, April 18-22, 1983, Paper No. 135. 10.35 Non-Class 1 Mechanical Implementation Guideline and Mechanical Tools, Revision 4, Electric Power Research Institute, Palo Alto, CA. 2006. Report 1010639. © Copyright 2023 by NuScale Power, LLC 237

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 10.36 Boric Acid Corrosion Guidebook, Managing Boric Acid Corrosion Issues at PWR Power Stations, Revision 1, Electric Power Research Institute, Palo Alto, CA. 2001. Report 1000975. 10.37 Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Storage and Transportation Applications, Revision 1, Electric Power Research Institute, Palo Alto, CA. 2022. Report 3002018496. 10.38 S. Lingham (NRC Project Manager) Letter to R. Bement (CNO Arizona Public Service Company), Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments to Revise Technical Specifications to Incorporate Updated Criticality Safety Analysis (CAC Nos. MF7138, MF7139, and MF7140), United States Nuclear Regulatory Commission, Rockville, MD, July 2017. Enclosure 4 contains the Safety Evaluation. ADAMS Accession No. ML17188A412. 10.39 Letter from Kristopher Cummings (NEI) to Brian J. Benney (United States Nuclear Regulatory Commission), Submittal of NEI 16-03, Guidance for Monitoring of Fixed Neutron Absorbers in Spent Fuel Pools, Revision 0, dated August 2016, May 26, 2017. ADAMS Accession No. ML17263A133. 10.40 American Society for the Testing of Materials ASTM G46-21, Standard Guide for Examination and Evaluation of Pitting Corrosion, American Society for the Testing of Materials, West Conshohocken, PA, August 2021. 10.41 United States Nuclear Regulatory Commission NUREG-0800 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 9.1.1, Revision 3, Criticality Safety of Fresh and Spent Fuel Storage and Handling, United States Nuclear Regulatory Commission, Rockville, MD, March 2007. ADAMS Accession No. ML070570006. © Copyright 2023 by NuScale Power, LLC 238

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Appendix A Level D Member Stresses Figure A-1 Corner Post Shear Stress [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure A-2 Outer Brace Shear Stress [

Figure A-3 Outer Fuel Tubes Shear Stress [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure A-4 Inner Fuel Tube Shear Stress [

Figure A-5 Bumper Web Shear Stress [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure A-6 Bumper Flange Shear Stress [

Figure A-7 Bottom Side Plate Shear Stress [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure A-8 Baseplate Shear Stress [

Figure A-9 Top Outer Grid Shear Stress [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure A-10 Top Inner Grid Shear Stress [

Figure A-11 Corner Post in Bottom and Top Grid Shear Stress [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure A-12 Bottom Outer Grid Shear Stress [

Figure A-13 Bottom Inner Grid Shear Stress [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Appendix B Criticality Benchmark Analysis B.1 Overview The software codes used for criticality analysis are validated against benchmark experiments. The validation determines the calculation bias and the uncertainty of the bias that is associated with the modeling methodology, the code, and the cross-section library used to perform the criticality analysis. Guidance for the validation is provided in NUREG/CR-6698 (Reference 10.23). Criticality codes are verified by comparing benchmark calculations to actual critical benchmark experiments. The difference between the calculated reactivity and the experimental reactivity is referred to as calculational bias. A trend analysis is performed to determine whether the calculational bias is a function of system parameters such as fuel lattice separation, fuel enrichment, neutron absorber properties, reflector properties, or fuel/moderator volume ratio. The purpose of this validation is to statistically determine the magnitude of the calculational bias and the uncertainty of the bias, and whether such dependencies exist so that they are properly accounted for in the criticality analysis. The parameter range for the selection of critical experiments for the benchmark cases is shown in Table B-1. Table B-1 Parameter Range for Critical Experiment Selection Parameter [ ] Fissionable Material [ ] Isotopic Composition [ ] Physical Form [ ] Temperature [ ] Moderator to Fuel Area Ratio [ ] Moderation Material [ ] Moderator Density [ ] Neutron Absorbing Material [ ] Geometry [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 B.2 Methodology B.2.1 Bias and Bias Uncertainty If no trends of statistical significance are found and the data are normally distributed, then the bias and uncertainty is based on a single-sided tolerance limit with the following equations. When comparing the experimentally measured k-effective (kexp) to the calculated k-effective (kcalc), the values are normalized as shown in NUREG/CR-6698, Equation 9 (Reference 10.23): k norm = k calc k exp Equation B-1 In addition, the errors are combined statistically as shown in NUREG/CR-6698, Equation 9: 2 2 t = + Equation B-2 calc exp The weighted mean value of k-effective is calculated with the following set of equations as shown in NUREG/CR-6698, Equations 4 through 7: Variance about the mean: 1 -*

                                           § -----------              1                                 2
                                           © n - 1¹          ----- 
                                                                        - ( k - k eff )

2 eff i 2 i s = ---------------------------------------------------------------- Equation B-3 1 1 n 2 i Average total uncertainty: 2 n

                                                     = -------------           -                                        Equation B-4 1

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Weighted mean keff: 1

                                                                -k 2 eff      i i

k eff = -----------------------

                                                                          -                                  Equation B-5 1

2 ------ i Square root of the pooled variance: 2 2 Sp = s + Equation B-6 Finally, the bias is determined as: Bias = k eff - 1 Equation B-7 if keff is less than one; otherwise bias = 0. And the uncertainty on the bias is: C x Sp Equation B-8 Where C is the 95/95 single-sided tolerance factor dependent on the sample size. Table 2.1 of NUREG/CR-6698 provides tolerance factors for sample sizes up to 50; Sandia Corporation Monograph SCR-607 (Reference 10.25) provides more expansive data. B.2.2 Normality The use of the pooled variance and the single sided tolerance limit is predicated on the requirement for the data to be representative of a normal distribution. In the details given by the example data of the NRC guidance, NUREG/CR-6698 (Reference 10.23), a Shapiro-Wilk test is used. While appropriate for that given example, the data set in validation has a much larger population. An expanded Shapiro-Wilk test for sample sizes between 12 and 2000 is used (Reference 10.25). Additionally, a more robust test, D from ANSI N15.15-1974 (Reference 10.26) is used to verify that the sampled experiments are representative of a normally distributed data set. B.2.3 Expanded Shapiro-Wilk Test For a sample size of n, sort the data in ascending order: x 1 x 2 ... x n-1 x n Equation B-9 © Copyright 2023 by NuScale Power, LLC B-3

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Define the values m1, ..., mn as: m i = NORM.S.INV ( ( i - 0.375 ) ( n + 0.25 ) ) Equation B-10 Where NORM.S.INV is the Excel function for the inverse of the standard normal cumulative distribution: Define m as: m = mi Equation B-11 Define u as: 1 u = ------- Equation B-12 n Define the coefficients a1, ..., an: 5 4 3 2 a n = -2.706056u + 4.434685u - 2.07119u - 0.147981u + 0.221157u + m n m Equation B-13 5 4 3 2 a n-1 = -3.582633u + 5.682633u - 1.752461u - 0.293762u + 0.042981u + m n m Equation B-14 m1 a i = ------- for 2 < i < n-1 Equation B-15 a 2 = -a n-1 and a 1 = -a n Equation B-16 where: 2 2 m-2xm -2xm n n-1

                                           = -----------------------------------------------------
                                                                                                  -                       Equation B-17 2                    2 1-2xa -2xa n                    n-1 The W statistic is defined by:

2

                                                           § a xx *
                                                           ©            i         1¹ W = --------------------------------                                       Equation B-18 2

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 For values of n between 12 and 5,000 the statistic ln(1-W) is approximately normally distributed with the following mean and standard deviation: 3 2

                 = 0.0038915 x ( ln ( n ) ) - 0.083751 x ( ln ( n ) ) - 0.31082 x ( ln ( n ) ) - 1.5861             Equation B-19 2

0.0030302 x ( ln ( n ) ) - 0.082676 x ln ( n )

                                  = e                                                                               Equation B-20 The z statistic is defined as:

ln ( 1 - W ) - z = ---------------------------------- Equation B-21 The z statistic is tested using the standard normal distribution. Finally, the p-value is computed using an Excel formula: p = 1 - NORM.S.DIST(z,TRUE) Equation B-22 If the p-value is the null hypothesis that the original data are normally distributed is rejected. For this application is 0.05. B.2.4 ANSI N15.15 Normality Test Evaluating the test statistic ANSI N15.15-1974, Section 7.2.4 (Reference 10.26): T D' = --- Equation B-23 S Where NUREG/CR-6698, Section 7.2.2 (Reference 10.23): n (n + 1) T = i - ----------------- x i 2 Equation B-24 i=1 And NUREG/CR-6698, Section 4.2.2: 2 2 2 ( xi ) S = x i n Equation B-25 Critical values for the 95th percentile would be P(0.025) and P(0.975) for the two-tailed test. © Copyright 2023 by NuScale Power, LLC B-5

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 B.2.5 Non-Parametric Method If the data used to determine the bias and uncertainty (Section B.2.1) are found not to be normal, or if statistically significant trends are found and the residuals are found not to be normal (Section B.2.7), then the non-parametric method must be used to determine the bias value. From NUREG/CR-6698, Section 2.4.4 (Reference 10.23), the following equation determines the confidence that a fraction of the population is above the mth lowest observed value: m-1 n! j (n - j)

                                     = 1-        j!---------------------

( n - j )!

                                                                         -(1 - q) q                             Equation B-26 j=0 where:

q = the desired population fraction (normally 0.95) n = the number of data in the sample m = the rank order index from the small sample (m=1) to the largest sample (m=n) For m=1 the equation reduces to: n n

                                             = 1 - q = 1 - 0.95                                                Equation B-27 For n=45 and m=1, =0.90 and the Non-Parametric Margin shown in Table 2.2 of NUREG/CR-6698 is not required. N must be 59 to obtain a 95% degree of confidence that 95% of the population is greater than the smallest observed value. For a sample size of 93, the second smallest value is appropriate. For a sample size of 124 the third smallest value is appropriate, and for a sample size of 153, the fourth smallest sample is appropriate while maintaining a confidence level of 95/95.

For a non-parametric analysis, KL if determined by: th th K L = m smallest k norm - t of m smallest k norm - Non-parametric Margin (NPM) Equation B-28 Where: m is a function of sample size as discussed above NPM is taken from Table 2.2 of NUREG/CR-6698 as required by the value of © Copyright 2023 by NuScale Power, LLC B-6

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 If the smallest value of knorm is > 1.0, then this equation is used: K L = 1 - S p - NPM Equation B-29 where Sp is the pooled variance as discussed in Section B.2.1. B.2.6 Trend Analysis Trending of knorm against a particular parameter (e.g., enrichment, fuel rod pitch, etc.) is performed with the weighted least squares fit method of NUREG/CR-6698 (Reference 10.23), equations 10 through 15. For Y(x) = a + bx Equation B-30 2

                                            §           x              y                   x               x i y i*¸ 1

a = --- ¨ -----i- -----i- - -----i- --------- Equation B-31

                                        ¨                 2              2                   2                  2¸
                                            ©                                                              ¹ i              i                   i                i
                                             §                        xi yi                    x               y

1 1-b = --- ¨ ----- --------- - -----i- -----i-¸ Equation B-32

                                        ¨                  2                2                     2              2¸
                                             ©                                                              ¹ i                i                     i               i 2                             2 x           §             x

1

                                        =              ------         -----i- - ¨             -----i-¸                     Equation B-33 2               2 ¨                    2¸
                                                                         ©                       ¹ i                i                       i 1
                                                      ---- ( xi - x ) ( yi - y )

i R = ------------------------------------------------------------------------------- - Equation B-34 1 2 1 2

                                                  ------ ( x i - x )

2 ------ ( y i - y ) 2 i i NUREG/CR-6698 omitted a subscript on y in the equations for a and b. The equations above have the subscript included. The standard error of the coefficients a and b are computed using equations for ordinary least squares fit: 2 SS x = ( xi - xmean ) Equation B-35 © Copyright 2023 by NuScale Power, LLC B-7

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 SS xy = ( ( xi - xmean ) x ( yi - ymean ) ) Equation B-36 2 SS y = ( yi - ymean ) Equation B-37 2 SS xy SS y - ------------ SS x S = ---------------------------- Equation B-38 n-1 2 x 1 mean Standard Error of a = S x --- + -------------- Equation B-39 n SS x Standard Error of b = ------------- S Equation B-40 SS x B.2.7 Determination of a Valid Trend Several factors are considered to determine if a trend is statistically significant The magnitude of the slope R2 which ranges from 0 (no correlation) to 1 (perfect correlation) F test T test If a correlation is deemed to be significant a single-sided tolerance band is evaluated to determine an appropriate KL. The residuals from the curve fit must be normal for the tolerance band to be valid. The F statistic is computed with the following equations: 2 Total Sum of Squares = SST = ( yi - ymean ) Equation B-41 2 Regression Sum of Squares = SS R = ( ( a + b x xi ) - yi ) Equation B-42 SS R v1 F statistic = -------------------- Equation B-43 SS T v2 © Copyright 2023 by NuScale Power, LLC B-8

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 where: v1 is always 1, v2 is n-2 (n is the number of observations), a and b are the coefficients from the weighted least square fit If the F statistic is greater than the critical value, then it is likely that the trend is meaningful. The critical value is calculated with the Excel function F.INV.RT(0.05, v1, v2), where 0.05 is the level of significance. The T statistic is computed as the slope of the trend divided by standard error of the slope. If the T statistic is greater than the critical value, then it is likely that the trend is significant. The critical value is computed with the Excel function T.INV.2T(0.05, v2), where 0.05 is the level of significance). B.3 Selection of Experiments A large set of experiments are selected for either the determination of the bias and bias uncertainty, or for the determination of trends, or for both uses. The complete list of experiments and the description of the experiments is provided in Table B-2. Table B-2 Critical Experiments Used for Bias Determination, Trending or Both [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table B-2 Critical Experiments Used for Bias Determination, Trending or Both (Continued) [

B.3.1 Selection of Experiments for Bias and Uncertainty, Enrichment Trend, Plutonium Trend (with all cases) The set of experiments that are used to determine the bias and the uncertainty on the bias are shown in Table B-3. This same selection is used for trending enrichment and Plutonium (with all cases). Table B-3 Benchmark Experiments Selected for Bias and Bias Uncertainty [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table B-3 Benchmark Experiments Selected for Bias and Bias Uncertainty (Continued) [

B.3.2 Selection of Experiments for Fuel Rod Pitch Trend The experiments shown in Table B-4 are chosen to sample the calculation of k effective with fuel rod pitch that is representative of the FSRs. The range of fuel rod pitch is expanded to [ ] The expanded area of applicability serves to improve the statistical significance of the trend. Table B-4 Benchmark Experiments Selected for Fuel Rod Pitch Trend [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table B-4 Benchmark Experiments Selected for Fuel Rod Pitch Trend (Continued) [

B.3.3 Selection of Experiments for Fuel Assembly Separation Trend The experiments shown in Table B-5 are chosen to sample the calculation of k-effective with FA separation that is representative of the FSRs. The FA separation ranged from [ ] Table B-5 Benchmark Experiments Selected for Fuel Assembly Separation Trend [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 B.3.4 Selection of Experiments for Soluble Boron Trend The experiments shown in Table B-6 are chosen to sample the calculation of k-effective with soluble boron concentration that is representative of the FSRs. The soluble boron concentration range is expanded to [ ] to adequately cover the expected range. Experiments that had a hexagonal fuel lattice are not included. Table B-6 Benchmark Experiments Selected for Soluble Boron Trend [

B.3.5 Selection of Experiments for Boron Separator Plate Areal Density Trend The experiments shown in Table B-7 are chosen to sample the calculation of k-effective with boron separator plates that are representative of the FSRs. The boron atom density ranged from [ ] This adequately covers the expected boron density for the FSRs, [ ] It is not possible to analyze the impact of boron areal density within the suggested +/-20% and the range of boron concentration in the separator plates is increased until a minimal © Copyright 2023 by NuScale Power, LLC B-15

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 sample size is obtained. Without this extension, there are only a total of five experiments in the range of the nominal boron density, which is far too small a sample to determine if a trending bias is present. There are Boroflex experiments that are not chosen, as this material is specifically excluded from the construction of the FSRs. Cases that use stainless steel separator plates are included, even when impregnated with boron, because it is expected that the construction makes use of large amounts of stainless steel, if not for the poison plate substrate, it still separates the FAs. Table B-7 Benchmark Experiments Selected for Separator Plate Boron Areal Density Trend [

B.3.6 Selection of Experiments for Moderator to Fuel Area Ratio Trend The experiments shown in Table B-8 are chosen to sample the calculation of k-effective with moderator to fuel ratios that are representative of the FSRs. The range of ratios are from [ . ] The range is extended to improve the statistical significance of the trend. Experiments with a hexagonal pitch are excluded. In most cases, the moderator to fuel ratio is not given in any reference, so a simple calculation is performed to compare the area of the fuel to the area of the moderator. Table B-8 Benchmark Experiments Selected for Moderator to Fuel Area Ratio Trend [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table B-8 Benchmark Experiments Selected for Moderator to Fuel Area Ratio Trend (Continued) [

B.3.7 Selection of Experiments for Neutron Spectrum The experiments shown in Table B-9 are chosen to sample the calculation of k-effective with neutron spectrum, as represented by the EALF, that are representative of the FSRs. The range of ratios are from [ ] This adequately covers the expected range for the FSRs of [

                              ] as shown in Table 6-34.

Table B-9 Benchmark Experiments for Neutron Spectrum Trend [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table B-9 Benchmark Experiments for Neutron Spectrum Trend (Continued) [

B.3.8 Selection of Experiments for Plutonium Trend The experiments shown in Table B-10 are chosen to show the relationship of the MOX experiments to the enriched uranium experiments. The experiments shown in Table B-10 show the selected MOX cases by themselves to better understand the behavior of 240Pu enrichment. Table B-10 Benchmark Experiments Selected for 240Pu Trend [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The applicant needs to include in the validation fresh UO2, MOX and HTC experiments. The code bias and uncertainty needs to be determined both with UO2 experiments alone and with HTC and MOX experiments included. The appropriate bias and uncertainty from each of these cases are included for fresh and spent fuel, respectively. However, the Haut Taux de Combustion (HTC) experiments are no longer readily available. An evaluation of the SCALE code system with and without the HTC experiments is provided in Table 6.27 of NUREG/CR-7109 (Reference 10.27). These results indicate an improvement of the code bias and uncertainty when the HTC experiment data are included. Based on the publicly available data it is reasonable to conclude that the HTC experiments do not increase the code bias or uncertainty. B.4 Results of Benchmark Calculations A summary of the pertinent parameters for each experiment is shown in Table B-11, along with the results of each KENO V.a case. © Copyright 2023 by NuScale Power, LLC B-19

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

Table B-11 Critical Experiment Parameters and KENO V.a Results [

  © Copyright 2023 by NuScale Power, LLC B-20

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

Table B-11 Critical Experiment Parameters and KENO V.a Results (Continued) [

              © Copyright 2023 by NuScale Power, LLC B-21

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

Table B-11 Critical Experiment Parameters and KENO V.a Results (Continued) [

              © Copyright 2023 by NuScale Power, LLC B-22

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

Table B-11 Critical Experiment Parameters and KENO V.a Results (Continued) [

              © Copyright 2023 by NuScale Power, LLC B-23

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

Table B-11 Critical Experiment Parameters and KENO V.a Results (Continued) [

              © Copyright 2023 by NuScale Power, LLC B-24

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

Table B-11 Critical Experiment Parameters and KENO V.a Results (Continued) [

              © Copyright 2023 by NuScale Power, LLC B-25

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

Table B-11 Critical Experiment Parameters and KENO V.a Results (Continued) [

              © Copyright 2023 by NuScale Power, LLC B-26

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 B.5 Trending Analysis A regression analysis is performed to evaluate potential biases that may trend with an independent variable. The following physical or spectral parameters are investigated: 235U enrichment Fuel rod pitch FA separation Boron concentration in moderator Boron areal density in separator plates Moderator to fuel area ratio Neutron spectrum as quantified by EALF 240Pu enrichment in just MOX cases 240Pu enrichment in MOX cases with UO2 cases included Each trend analysis uses a different set of experiments as described in Section B.3.1 through Section B.3.8. The results of the regression analysis are shown in Table B-12. The results for each trend analysis are shown in three plots: A plot of knorm versus the parameter of interest, which includes error bars that represent the value of t, also shows the weight linear curve fit and the tolerance band. The tolerance band is always shown even if it is not used. A histogram of the curve fit residuals A normality plot of the curve fit residuals © Copyright 2023 by NuScale Power, LLC B-27

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0

Table B-12 Regression Analysis for Possible Bias Trending Variables [

  © Copyright 2023 by NuScale Power, LLC B-28

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 The plots shown in Figure B-1 through Figure B-27 illustrate the data that the regression analysis made use of. The error bars plotted show the total uncertainty in the estimate, combining both the experimental and calculated uncertainty as shown in Table B-11. The regression analysis summarized in Table B-12 shows that there is a possible trend for Enrichment, Assembly Separation, and Neutron Spectrum (EALF). However, the curve fit residuals are not normal for all three, so the tolerance band is not valid. This result forces the use of the non-parametric method (Section B.2.5) over the use of the single-sided tolerance limit (Section B.2.1). Figure B-1 Regression Analysis of 235U Enrichment [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-2 Histogram of Curve Fit Residuals for 235U Enrichment [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-3 Normality Plot of Curve Fit Residuals for 235U Enrichment [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-4 Regression Analysis of Fuel Rod Pitch [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-5 Histogram of Curve Fit Residuals for Fuel Rod Pitch [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-6 Normality Plot of Curve Fit for Residuals for Fuel Rod Pitch [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-7 Regression Analysis of Fuel Assembly Separation [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-8 Histogram of Curve Fit Residuals for Fuel Assembly Separation [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-9 Normality Plot of Curve Fit Residuals for Fuel Assembly Separation [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-10 Regression Analysis of Dissolved Boron Concentration [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-11 Histogram of Curve Fit Residuals for Dissolved Boron Concentration [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-12 Normality Plot of Curve Fit Residuals for Dissolved Boron Concentration [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-13 Regression Analysis of Separator Plate Areal 10B Density [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-14 Histogram of Curve Fit Residuals for Separator Plate Area 10B Density [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-15 Normality Plot of Curve Fit Residuals for Separator Plate Area 10B Density [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-16 Regression Analysis of Moderator to Fuel Area Ratio [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-17 Histogram of Curve Fit Residuals for Moderator to Fuel Area Ratio [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-18 Normality Plot of Curve Fit Residuals for Moderator to Fuel Area Ratio [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-19 Regression Analysis of Neutron Spectrum [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-20 Histogram of Curve Fit Residuals for Neutron Spectrum [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-21 Normality Plot of Curve Fit for Neutron Spectrum [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-22 Regression Analysis of 240Pu Enrichment (All Cases) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-23 Histogram of Curve Fit Residuals for 240Pu Enrichment (All Experiments) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-24 Normality Plot of Curve Fit Residuals for 240Pu (All Experiments) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-25 Regression Analysis of 240Pu Enrichment (MOX Experiments) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-26 Histogram of Curve Fit Residuals of 240Pu Enrichment (Mixed Oxide Fuel Experiments) [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Figure B-27 Normality Plot of Curve Fit Residuals for 240Pu Enrichment (Mixed Oxide fuel experiments) [

B.6 Test for Normality Table B-13 summarizes the data used to calculate the bias and the bias uncertainty. The cases are those listed in Table B-2. knorm and t are calculated as described in Section B.2.1, using the experimental and calculated results shown in Table B-11. Using the D test as described in Section B.2.4, the test statistic is found to be: [

The critical values for a sample of [ ] from ANSI N15.15-74 (Reference 10.22). Because the calculated value for D is not within the critical values for a 95% probability, the normality hypothesis must be rejected. The Shapiro-Wilk test described in Section B.2.3 also rejects the normality hypothesis. © Copyright 2023 by NuScale Power, LLC B-55

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table B-13 Data Used for Bias and Bias Uncertainty [

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 Table B-13 Data Used for Bias and Bias Uncertainty (Continued) [

B.7 Bias and Bias Uncertainty The distribution of knorm did not pass the normality tests and the curve fit residuals for potentially significant trends also did not pass the normality tests so the bias must be computed with the non-parametric method as discussed in Section B.2.5. For 153 experiments and m=4, using the equations provided in Section B.2.5 =0.9506. The actual number of experiments used is [ ] so the actual value of is even higher. Since > 0.90 the Non-Parametric Margin may be set to zero. From Table B-11 the fourth smallest value of knorm is associated with experiment [ ] Using the equation of KL from Section B.2.5 and data from Table B-11 for experiment [ ] [

This equation gives a bias at the 95/95 level of: [

As the value of Bias is at the 95/95 level, no uncertainty value is needed. B.8 Benchmark Summary A total of [ ] benchmark critical experiments are used to establish the bias for use in the criticality safety analysis of the FSRs. The trending analyses demonstrate that statistically significant trends exist for Enrichment, Assembly Separation, and Neutron Spectrum (EALF). However, the curve fit residuals are not normal for any of these trends, so the tolerance band is not valid. The distribution of the [ ] values of knorm is also not normal. For these two reasons the bias is determined by the non-parametric method at a value of [ ] © Copyright 2023 by NuScale Power, LLC B-61

NuScale US460 Fuel Storage Rack Design Topical Report TR-145417-NP Revision 0 B.9 Implementation / Use The results presented in Section B.8 are applicable to the FSRs over the area of applicability given in Table B-14. The area of applicability for enrichment is [ ] 235U, which covers the limiting enrichment of [ ] 235U for the analysis. The area of applicability for neutron spectrum, as measured by EALF, is [

Table B-14 Area of Applicability for Bias and Bias Uncertainty Parameter Range 235U Enrichment [ ] Fuel Rod Pitch [ ] Fuel Assembly Separation [ ] Soluble Boron Concentration [ ] Area Density of Boron-10 in Separator Plates [ ] Moderator to Fuel Ratio [ ] Neutron Spectrum, Energy of Average Lethargy of Fission [ ] © Copyright 2023 by NuScale Power, LLC B-62

LO-151775 : Affidavit of Carrie Fosaaen, AF-151777 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Power, LLC AFFIDAVIT of Carrie Fosaaen I, Carrie Fosaaen, state as follows: (1) I am the Vice President of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying report reveals distinguishing aspects about the process by which NuScale develops its statistical subchannel analysis methodology. NuScale has performed significant research and evaluation to develop a basis for this fuel rack design and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed report entitled NuScale US460 Fuel Storage Rack Design Topical Report, TR-145417-P, Revision 0. The enclosure contains the designation Proprietary at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, (( }} in the document. AF-151777 Page 1 of 2

(5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury thatat the foregoing is true and correct. co Executed on October , 2023. Carrie Fosaaen AF-151777 Page 2 of 2

LO-151775 : Affidavit of Morris Byram, Framatome NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360.0500 Fax 541.207.3928 www.nuscalepower.com

AFFIDAVIT

1. My name is Morris Byram. I am Manager, Licensing & Regulatory Affairs for Framatome Inc. (Framatome) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.

3. I am familiar with the Framatome information contained in enclosure 1 to NuScale letter LO-151775 dated October 9, 2023, entitled NuScale US460 Fuel Storage Rack DesignTopical Report, TR-145417-P, Revision 0, referred to herein as Document.

Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.

6. The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:

(a) The information reveals details of Framatomes research and development plans and programs or their results. (b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service. (c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome. (d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability. (e) The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome. The information in this Document is considered proprietary for the reasons set forth in paragraph 6(c) and 6(d) above.

7. In accordance with Framatomes policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

I declare under penalty of perjury that the foregoing is true and correct. Executed on: (10/10/2023) BYRAM Morris Digitally signed by BYRAM Morris Date: 2023.10.10 08:09:35 -07'00' (NAME) morris.byram@framatome.com}}

v t e Letter Sequence Request
EPID:L-2023-TOP-0052, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML23339A003, ML23339A004
MONTHYEAR2023-12-04 [Table View]

ML23283A336

Text

SUBJECT:

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