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LLC, Revision 1 to Standard Design Approval Application, Part 2, Chapter 6, TR-123952-NP, Containment Leakage Integrity Assurance
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NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Licensing Technical Report NuScale Containment Leakage Integrity Assurance December 2022 Revision 0 Docket: 52-050 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Licensing Technical Report Revision History Revision Date Notes 0 12/2022 Initial Issuance

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Licensing Technical 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.

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NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Licensing Technical Report Department of Energy Acknowledgement 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.

NuScale Containment Leakage Integrity Assurance TR-123952-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 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.4 Containment Leakage Integrity Assurance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1.5 Regulatory Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 1.6 Abbreviations and Definitions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 1.7 Containment Leakage Integrity Assurance Overview . . . . . . . . . . . . . . . . . . . . . . . . . . 10 2.0 NuScale Containment Vessel Structure. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 2.1 NuScale Containment Vessel Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 2.2 Type B Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 2.3 Type C Testing. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 2.4 Containment Overall Leakage Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.0 NuScale Containment System Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 3.1 Containment Vessel Design. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.1.1 Preservice Tests and Inspections (Containment Vessel). . . . . . . . . . . . . . . . . . 19 3.2 Preservice Tests and Inspections (Type B and Type C Components) . . . . . . . . . . . . . 19 3.2.1 Type B Penetrations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.2.2 Access Ports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 3.2.3 Electrical Penetration Assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 3.2.4 Instrument Seal Assemblies. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.2.5 Emergency Core Cooling System Trip and Reset Body-to-Bonnet Seals . . . . . 23 3.2.6 Containment Vessel Closure Flange . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3.3 Type C Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 3.3.1 Primary System Containment Isolation Valves . . . . . . . . . . . . . . . . . . . . . . . . . 28 3.3.2 Secondary System Containment Isolation Valves . . . . . . . . . . . . . . . . . . . . . . . 29 3.3.3 Steam Generator Thermal Relief Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 3.3.4 Main Steam Isolation Valves and Bypass Valves . . . . . . . . . . . . . . . . . . . . . . . 31 3.3.5 Feedwater Plenum Cover Drain Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 4.0 Inservice Inspection and Testing of Containment . . . . . . . . . . . . . . . . . . . . . . . . . . 32

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table of Contents 4.1 Inservice Inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.1.1 Weld Inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.1.2 Visual Inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.1.3 Bolting Inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4.1.4 Steam Generator Inspections and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.1.5 Type B Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 4.2 Inservice Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 5.0 Type B Local Leak Rate Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 5.1 Type B Test Method. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 5.2 Electrical Penetration Assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 5.3 Instrument Seal Assemblies. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 5.4 Access Ports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 5.5 Emergency Core Cooling System Pilot Valve Bodies . . . . . . . . . . . . . . . . . . . . . . . . . . 35 5.6 Containment Vessel Closure Flange . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 5.7 2Bolting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 6.0 Type C Local Leak Rate Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 6.1 Type C Test Method. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 7.0 Containment Leakage Rate Test Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 7.1 Challenges Associated with Type A Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 7.2 Containment Leakage Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 7.3 Test Frequency . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 7.4 Test Results and Reporting Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 7.5 Special Testing Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 7.5.1 As-Found Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 7.5.2 As-Left Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 7.5.3 Preconditioning. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 7.5.4 Reverse Direction Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 7.5.5 Modifications After Preoperational Testing. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 7.5.6 Option B Performance-Based Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 8.0 Material Selection and Aging Degradation Leakage Rate Test Program . . . . . . . . 44 8.1 Material Selection and Operating Conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 8.1.1 Pool Water Chemistry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table of Contents 8.1.2 Reactor Coolant System Coolant Chemistry . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 8.2 Aging Degradation Assessment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 8.2.1 Fatigue . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 8.2.2 Boric Acid Corrosion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 8.2.3 Primary Water Stress Corrosion-Cracking . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48 8.2.4 Stress Corrosion-Cracking of Austenitic Stainless Steels . . . . . . . . . . . . . . . . . 48 8.2.5 Stress Corrosion-Cracking of Pressure-Retaining Bolting Materials . . . . . . . . . 49 8.2.6 Irradiation Embrittlement of Lower Containment Vessel . . . . . . . . . . . . . . . . . . 50 8.2.7 Stress Corrosion-Cracking of F6NM Martensitic Stainless Steel . . . . . . . . . . . . 52 9.0 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 Appendix A Containment Isolation Summary Figures . . . . . . . . . . . . . . . . . . . . . . . . . . .A-1 Appendix B Type B Containment Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-1 Appendix C List of Type C Containment Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . .C-1 Appendix D Type A Testing Challenges . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-1 D.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-1 D.2 Temperature. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-1 D.3 Temperature Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-2 D.4 Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-3 D.5 Leak Rate Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-4 D.6 Alternate Testing Arrangements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-5 D.7 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-5

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 List of Tables Table 1-1 Containment Leakage Integrity Program Elements . . . . . . . . . . . . . . . . . . . . . . . 4 Table 1-2 Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 Table 1-3 Definitions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 Table 2-1 NuScale Containment Leak Rate Test Comparison. . . . . . . . . . . . . . . . . . . . . . 14 Table 2-2 Maximum Allowable Containment Leakage Rate Limits . . . . . . . . . . . . . . . . . . 16 Table 8-1 Containment Vessel Pressure-Retaining Materials . . . . . . . . . . . . . . . . . . . . . . 45 Table 8-2 Target Limits for Reactor Pool Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 Table 8-3 Reactor Coolant System Coolant Chemistry . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 Table A-1 Simplified Figures Illustrating the Containment Pressure Boundary for the General Design Criteria 55, General Design Criteria 56 and General Design Criteria 57 Piping Systems of the Containment Vessel.. . . . . . . . . . . . .A-1 Table B-1 Type B Containment Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-2 Table C-1 Penetration CNV1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-1 Table C-2 Penetration CNV2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-1 Table C-3 Penetration CNV3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-3 Table C-4 Penetration CNV4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-4 Table C-5 Penetration CNV5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-5 Table C-6 Penetration CNV6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-6 Table C-7 Penetration CNV7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-7 Table C-8 Penetration CNV10 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-8 Table C-9 Penetration CNV11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-9 Table C-10 Penetration CNV12 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-10 Table C-11 Penetration CNV13 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-11 Table C-12 Penetration CNV14 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-12 Table C-13 Penetrations CNV22 and CNV23 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C-13

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 List of Figures Figure 3-1 Containment Vessel Head . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 Figure 3-2 Containment Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 Figure 3-3 Containment Vessel Head / Control Rod Drive Mechanism Access Flange . . . 22 Figure 3-4 Electrical Penetration Assembly Modules (Typical) . . . . . . . . . . . . . . . . . . . . . . 23 Figure 3-5 Emergency Core Cooling System Valve Trip/Reset Pilot Assembly (Simplified Diagram). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 Figure 3-6 Containment Vessel Closure Flange Seal and Test Port . . . . . . . . . . . . . . . . . . 25 Figure 3-7 Primary System Containment Isolation Valve and Secondary System Containment Isolation Valve Arrangement on Containment Vessel Head . . . . . 27 Figure 3-8 Primary System Containment Isolation Valve Dual Valve, Single Body Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 Figure 3-9 Feedwater Isolation Valve Dual Valve, Single Body Design . . . . . . . . . . . . . . . 30 Figure 3-10 CombinedMain Steam Isolation Valve and Main Steam Isolation Bypass Valve, Single Unit Design. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 Figure 8-1 Lower Containment Vessel Pressure-Retaining Materials . . . . . . . . . . . . . . . . . 52 Figure 8-2 Stress Corrosion Cracking Depth as a Function of Hardness, Martensitic Stainless Steels . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 Figure A-1 Containment System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-2 Figure A-2 Decay Heat Removal System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-3 Figure A-3 Steam Generator System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A-4 Figure D-1 Reactor Pressure Vessel, Containment Vessel, and Ultimate Heat Sink Temperature Gradients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .D-2

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Abstract This technical report describes the NuScale Power, LLC (NuScale) Containment Leakage Integrity Program (CLIP). This program provides assurance that leakage integrity of containment is maintained and that containment leakage does not exceed allowable leakage rate values. The CLIP is a consolidation of programs described in the US460 Standard Design Final Safety Analysis Report (FSAR). CLIP elements are implemented under other programs as described in this report and the FSAR. The requirements of Title 10 of the Code of Federal Regulations (CFR)

Part 50, Appendix A, General Design Criterion 52 (GDC 52) state that containments shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure. The requirements of 10 CFR 50, Appendix J, Type A tests, include test specifications directly related to GDC 52 design requirements. The CLIP integrates containment vessel (CNV) flange design, which remains sealed at design pressure.

preservice leak test at design pressure performed for all CNVs.

initial (first-of-a-kind) containment vessel preservice leak test at design pressure performed with the vessel fully assembled with flanges in place.

preservice 10 CFR 50, Appendix J, Type B testing.

preservice 10 CFR 50, Appendix J, Type C testing.

post-installation and repair inspection and testing.

inservice inspection and examination.

periodic 10 CFR 50, Appendix J, Type B testing.

periodic 10 CFR 50, Appendix J, Type C testing.

This report provides relevant details of the CNV and containment systems (CNTS) designs, which support the CLIP in assuring containment leakage integrity. The CLIP provides leakage integrity assurance equivalent to the containment leakage testing requirements of 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.

This report provides supplemental information designed to inform the Nuclear Regulatory Commission evaluation of NuScale Final Safety Analysis Report, Section 6.2.6.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Executive Summary This technical report describes NuScales Containment Leakage Integrity Program (CLIP). The CLIP, supported by the NuScale containment vessel (CNV) and containment system (CNTS) design, provides leakage integrity assurance for the NuScale containment. As discussed in the NuScale US460 Standard Design Final Safety Analysis Report (FSAR), Section 6.2.6, the design supports an exemption from the requirements of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix A, General Design Criterion (GDC) 52 and 10 CFR 50, Appendix J, which specify the design for and performance of preoperational and periodic integrated leak rate testing at containment design pressure.

The CLIP, supported by the design and analysis of the CNV and CNTS, provides leakage integrity assurance for the containment. The CLIP is a consolidation of programs described in the FSAR. CLIP elements are implemented under other programs as described in this report and the US460 design. Each element of the CLIP is consistent with a corresponding element of an approved program for reactor pressure vessels or large light water reactor containments. The primary CLIP elements that provide leakage integrity assurance are containment vessel flanges that are designed to remain sealed at design pressure.

factory inspection and testing, including preservice leak testing at design pressure with zero visible leakage, to ensure initial containment leakage integrity in accordance with Inspections, Tests, Analyses, and Acceptance Criteria.

preservice and periodic Type B and C testing to ensure overall containment leakage does not exceed allowable leakage rate values (i.e., to quantify overall containment leak rates).

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC)Section III (Reference 9.1), and ASME Operation and Maintenance (OM) and ASME Section XI to ensure continued leakage integrity (i.e., ensures no unknown leak pathways develop over time).

Type B and Type C testing, inspections, and administrative controls (e.g., configuration management and procedural requirements for system restoration) to ensure leakage integrity associated with activity-based failure mechanisms (i.e., ensures CNV flanges and containment isolation valves [CIVs] remain within allowable leakage rate values after system and component modifications or maintenance).

While the CLIP described in this report does not conform to GDC 52 or Type A testing requirements, the advanced design and CLIP provide more comprehensive leakage integrity assurance than was considered when the subject regulations were adopted. This report provides a detailed overview of the key aspects of the testing, inspection, and design that ensures containment leakage integrity is maintained, including:

the overall containment leakage rate testing program, including the scope of the Type B and Type C testing to ensure adequate margin against design-basis leak rates Type B testing adequacy as ensured by the following.

- Containment vessel flanges are designed to remain sealed at design pressure.

- Preservice design pressure leakage test of the CNV with bolted flanges in place using as-designed flange covers installed with the design bolting materials, design bolting

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 assembly preloads, and design seals installed to demonstrate no visible leakage at design pressure. The test is performed at design pressure and a minimum temperature of 70 degrees F and a maximum temperature of 140 degrees F to minimize the possibility of brittle fracture.

The upper and lower halves of the CNV are assembled for the test of the first NuScale Power Module (NPM) of the initial plant.

After successful testing, the upper and lower halves of other CNVs may be tested separately.

- Covers with electrical and instrumentation penetrations may be substituted with blank covers having the same sealing design.

- Flange assembly uses positive verification to ensure proper flange loading from each stud.

- The test configuration may use blanked-off pipe ends in place of the CIVs.

- The acceptance criterion is no observed leakage from seals at examination pressure.

- The emergency core cooling system trip valve and reset valve body-to-bonnet joint seals are not considered to be a flanged connection and are not included in the containment flange bolting calculation or preservice design pressure leakage test.

the CNTS design as it applies to the containment function the ISI Program as it applies to the CNV the IST Program as it applies to CIVs materials selection and aging degradation assessment As described in this report, the containment design and CLIP ensure leakage integrity of containment is maintained and containment leakage does not exceed allowable leakage rate values. This report provides supplemental information designed to inform the Nuclear Regulatory Commission evaluation of NuScale Final Safety Analysis Report, Section 6.2.6 and SDAA, Part 7.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 1.0 Introduction 1.1 Purpose The purpose of this technical report is to describe the Containment Leakage Integrity Program (CLIP) as well as the containment vessel (CNV) and containment system (CNTS) design elements that ensure leakage integrity. This report evaluates the plant design and the CLIP against the requirements in Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix J as incorporated in DesignSpecific Review Standard, Section 6.2.6. This evaluation includes an assessment of the capability of the containment design to meet specific testing requirements in 10 CFR 50, Appendix J. This report identifies Type A requirements that are not to be applied because certain functions are achieved differently compared to existing large light water reactors (LLWRs) for which 10 CFR 50, Appendix J requirements were developed and written.

This report describes the approach to Type B and Type C testing through an evaluation of the containment design. This report provides supplemental information designed to inform the Nuclear Regulatory Commission evaluation of NuScale Final Safety Analysis Report (FSAR) Section 6.2.6 . As shown in the table below, each element of CLIP is consistent with a corresponding element of an approved program for reactor pressure vessels (RPVs) or LLWR containments, which have been incorporated within the FSAR.

This report provides a consolidated description of inspection, testing, and examination elements from several programs described in the FSAR related to containment leakage integrity. This report does not describe any elements that are not described in the FSAR.

Table 1-1 Containment Leakage Integrity Program Elements CLIP Element Licensing Requirement CNV flange design FSAR COL Item 6.2-2 Preservice inspection (Section 4.0) ASME BPVC ClassSection III (FSAR 6.2.6)

Fabrication structural integrity testing (Section 4.0) ASME BPVC ClassSection III (FSAR 6.2.6)

Preservice leakage testing FSAR 6.2.6, Chapter 14 Preservice Type B and Type C local leak rate test (LLRT) Technical Specifications (Section 5.5.9)

(Section 4.0)

Preservice Type B and Type C LLRT (Section 4.0) Initial Test Program (FSAR Table 14.2-43)

Post-installation/repair inspection & testing (Section 5.0) ASME BPVC ClassSection III / XI (FSAR 6.2.6)

Post-installation/repair inspection & testing (Section 5.0) TS (, Section 5.5.9)

Inservice inspection (ISI) and examination (Section 5.0) ASME BPVC ClassSection XI (FSAR 6.2)

Periodic Type B and Type C LLRT (Section 5.0) TS (, Section 5.5.9) 1.2 Scope This report describes the CLIP for the design and evaluates the CLIP against 10 CFR 50, Appendix J. This report describes the overall containment leakage rate testing (CLRT) program, including the scope and frequency of Type B and Type C testing of CNV penetrations.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 the CNTS design as it applies to CNV design and the containment function.

materials selection and aging degradation as it applies to the containment pressure boundary.

the ISI Program as it applies to the CNV.

the inservice testing (IST) program as it applies containment isolation valves (CIVs).

Type A integrated leak rate testing (ILRT) challenges.

1.3 Background

The design supports an exemption from GDC 52 to design the containment for ILRT and supports a licensees exemption from 10 CFR 50, Appendix J, for the performance of Type A preoperational and periodic integrated leakage rate testing.

This technical report describes the containment testing, inspection, and design criteria that ensure leakage integrity of containment is maintained and containment leakage does not exceed allowable leakage rate values, thereby satisfying the underlying purpose of GDC 52 and Type A testing.

1.4 Containment Leakage Integrity Assurance The CLIP provides containment leakage integrity by demonstrating the containment design can use LLRT to adequately ensure containment leakage integrity.

- Containment vessel flanges are designed to remain sealed at design pressure.

- Preservice design pressure leakage test of the CNV with CNV bolted flanges in place utilizing as-designed flange covers installed with the design bolting materials, design bolting assembly preloads, and design seals installed to demonstrate no leakage at design pressure.

The upper and lower halves of the CNV are assembled for the test of the first NPM of the initial NuScale plant.

After successful testing, the upper and lower halves of other CNVs may be tested separately.

Covers with electrical and instrumentation penetrations may be substituted with blank covers having the same sealing design.

- Flange assembly requires positive verification to ensure proper flange loading from each stud.

ensuring no unknown leakage pathways exist.

quantifying overall containment leak rates by LLRTs that provide accurate results for every potential leak path.

ensuring no unknown leak paths develop over time due to degradation.

ensuring no unknown leak paths develop due to activity-based failure mechanisms.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 1.5 Regulatory Requirements 10 CFR 52.137(a) states in part:

The application must contain a final safety analysis report that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole, and must include the following information:

(3) The design of the facility including:

(i) The principal design criteria for the facility. Appendix A to 10 CFR part 50, general design criteria (GDC), establishes minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have previously been issued by the Commission and provides guidance to applicants in establishing principal design criteria for other types of nuclear power units; (ii) The design bases and the relation of the design bases to the principal design criteria The introduction to 10 CFR 50, Appendix A, states in part:

Also, there may be water-cooled nuclear power units for which fulfillment of some of the General Design Criteria may not be necessary or appropriate. For plants such as these, departures from the General Design Criteria must be identified and justified.

10 CFR 50, Appendix A, GDC 52 states:

Criterion 52 - Capability for containment leakage rate testing. The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at containment design pressure.

10 CFR 50.54(o) states in part:

Primary reactor containments for water cooled power reactors shall be subject to the requirements set forth in appendix J to this part.

Appendix J to 10 CFR 50, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, states, in part:

One of the conditions of all operating licenses under this part and combined licenses under part 52 of this chapter for water-cooled power reactors as specified in 50.54(o) is that primary reactor containments shall meet the containment leakage test requirements set forth in this appendix. These test requirements provide for preoperational and periodic verification by tests of the leak-tight integrity of the primary reactor containment, and systems and components which penetrate

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 containment of water-cooled power reactors, and establish the acceptance criteria for these tests. The purposes of the tests are to assure that (a) leakage through the primary reactor containment and systems and components penetrating primary containment shall not exceed allowable leakage rate values as specified in the technical specifications or associated bases; and (b) periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of the containment, and systems and components penetrating primary containment. These test requirements may also be used for guidance in establishing appropriate containment leakage test requirements in technical specifications or associated bases for other types of nuclear power reactors.

1.6 Abbreviations and Definitions Table 1-2 Abbreviations Term Definition ANS American Nuclear Society ANSI American National Standard Institute ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials BPVC Boiler and Pressure Vessel Code CES containment evacuation system CFR Code of Federal Regulations CITF containment isolation test fixture CIV containment isolation valve CLIP Containment Leakage Integrity Program CLRT containment leakage rate testing CNTS containment system CNV containment vessel CRDM control rod drive mechanism CRDS control rod drive system CVCS chemical and volume control system DHRS decay heat removal system ECCS emergency core cooling system EFPY effective full-power year EPA electrical penetration assembly EPRI Electric Power Research Institute FSAR Final Safety Analysis Report FWIV feedwater isolation valve FWS feedwater system GDC general design criteria HRC Rockwell C hardness number HV Vickers hardness number I&C instrumentation and controls ICI incore instrumentation IGSCC intergranular stress corrosion-cracking ILRT integrated leak rate testing

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table 1-2 Abbreviations (Continued)

Term Definition ISA instrument seal assembly ISI inservice inspection IST inservice testing ITAAC Inspections, Tests, Analyses, and Acceptance Criteria LLRT local leak rate test LLWR large light water reactor MeV million electron volt MSIBV main steam isolation bypass valve MSIV main steam isolation valve MSS main steam system NEI Nuclear Energy Institute NPM NuScale Power Module NPS nominal pipe size (ASME B36.10M)

NRC Nuclear Regulatory Commission PSCIV primary system containment isolation valve PWHT post weld heat treatment PWR pressurized water reactor PWSCC primary water stress corrosion-cracking PZR pressurizer RAI request for additional information RCCWS reactor component cooling water system RCPB reactor coolant pressure boundary RCS reactor coolant system RPV reactor pressure vessel RVV reactor vent valve SCC stress corrosion-cracking scfh standard cubic foot per hour scfm standard cubic foot per minute SDAA Standard Design Approval Application SG steam generator SGS steam generator system SSC structures, systems, and components SSCIV secondary system containment isolation valve TS technical specification UHS ultimate heat sink Table 1-3 Definitions Term Definition GDC 55 penetration Reactor coolant pressure boundary penetrating containment. This type of penetration requires two NRC Quality Group A, ASME BPVC Class1 CIVs at each penetration.

GDC 56 penetration Containment boundary. This type of penetration requires two NRC Quality Group A, CIVs at each penetration.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table 1-3 Definitions (Continued)

Term Definition GDC 57 penetration Closed system lines that penetrate reactor containment and are neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere. This type of penetration requires one NRC Quality Group B, CIV at each penetration, or a second closed loop.

La Maximum allowable containment leakage rate at pressure Pa Pa Peak CNV accident pressure

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 1.7 Containment Leakage Integrity Assurance Overview The CLIP testing, inspection, and examination, as supported by the design and analysis of the CNV and CNTS, ensure leakage integrity is maintained for the containment. The CLRT, in combination with other CLIP elements, verifies the leakage integrity of the reactor containment by testing that the actual containment leakage rates do not exceed the values assumed in the applicable safety analysis calculations for design basis events.

The preoperational and periodic CLRT requirements and acceptance criteria that demonstrate leakage integrity of the CNTS and associated components are performed in accordance with 10 CFR 50, Appendix J and implemented through the licensees CLRT program described in Section 5.5.9 of the TS. The maximum allowable containment leakage rate is referred to as La, which corresponds to the peak accident pressure inside containment (Pa) (these terms are defined in 10 CFR 50, Appendix J). The containment penetrations and containment isolation barriers are designed to permit the periodic leakage testing described in GDC 53 and GDC 54 to verify leakage through the containment penetrations does not exceed the allowable leakage rate.

The design of the containment penetrations support performance of Type B and Type C testing in accordance with the guidance provided in American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8 (Reference 9.4) and Nuclear Energy Institute (NEI) 94-01 Rev. 3-A (Reference 9.6). The CNTS design accommodates the test method frequencies permitted by 10 CFR 50, Appendix J; Option B, Performance-Based Requirements. Applicants that reference the NuScale power plant US460 standard design approval will develop a Containment Leakage Rate Testing Program that will identify Option B to be implemented under 10 CFR 50, Appendix J. It is expected that the licensee will ultimately adopt Option B extended interval testing pursuant to the guidance in NEI 94-01 Rev. 3-A. Considerations for Option B extended interval testing are outlined in Section 7.5.

The CNTS is designed for flanged joints to remain sealed at design pressure. The containment is initially inspected and tested at the factory, including American Society of Mechanical Engineers (ASME) BPVC Section III hydrostatic testing with an acceptance criterion of zero visible leakage, to verify no unknown leak pathways exist. Additionally, a CNV preservice design pressure leakage test is performed that loads CNV bolted flange connections to containment design pressure and confirms no observed leakage under these conditions. Because potential leakage pathways are known and testable, preservice and periodic Type B and C testing quantify the overall containment leakage rate to verify maximum allowable leakage is not exceeded. Thus, the design and configuration of potential leak pathways, including CNV flanges and CIVs, accommodate LLRT requirements and acceptance criteria. Periodic inspection and testing verifies no unknown leakage pathways develop over time. Thus, any potential through-wall degradation is be precluded as a credible mechanism for containment leakage.

Post-maintenance inspection and testing, including Type B and C testing and administrative controls, verify that no unknown leakage pathways develop due to activity-based failure mechanisms during maintenance or modifications.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 2.0 NuScale Containment Vessel Structure The CNV design ensures leakage integrity through design, inspection and testing, other than ILRT, as required by GDC 52 and 10 CFR 50, Appendix J. NEI 94-01 Rev. 3-A describes the purpose of 10 CFR 50, Appendix J, for traditional large containment structures:

The purpose of Type A testing is to verify the leakage integrity of the containment structure. The primary performance objective of the Type A test is not to quantify an overall containment system leakage rate. The Type A testing methodology as described in ANSI/ANS-56.8-2002 serves to ensure continued leakage integrity of the containment structure. Type B and Type C testing ensures individual leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths.

Continued leakage integrity of the CNV structure is ensured by precluding through-wall degradation as a credible leakage mechanism. The CNV is a welded-metal vessel design, in contrast to existing pressurized water reactors (PWRs) that incorporate large containment building structures. The containment is designed for flanged joints to remain sealed at design pressure. Manufacturing acceptance tests and inspections are similar to RPV tests and inspections and are performed in a factory environment. Comprehensive ISI applying ASME Boiler and Pressure Vessel Code (BPVC) Class 1 criteria also ensures no new leakage paths develop over the life of the plant due to degradation.

Surface areas and welds are accessible for inspection. Additionally, a separate preservice design pressure leakage test is required for CNVs with CNV bolted flange connections in place to demonstrate no observed leakage using as-designed flange covers installed with the design bolting materials, design bolting assembly preloads, and design seals. This leakage test is required by an Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC). The first CNV of the initial plant shall be tested with the upper and lower halves of the CNV assembled. Penetration pathways are tested to Type B or Type C criteria at peak containment accident pressure. These features ensure continued leakage integrity of the CNTS is maintained without the need for Type A testing.

The NuScale CNV design is different from traditional containments in several fundamental aspects. These design differences provide reasons that the design does not need to conform with GDC 52 and 10 CFR 50, Appendix J, and necessitate alternative means of ensuring the leakage integrity of the NuScale containment. The major containment functional differences are:

The CNV is a high-pressure vessel with no internal subcompartments, classified as an ASME BPVC Class MC component, and constructed to ASME Code Class 1 vessel rules, constructed of stainless materials.

Penetrations are ASME Code Class 1 flanged joints capable of Type B testing, ASME Code Class 1 welded nozzles with isolation valves capable of Type C testing, or part of a closed system inside containment. Flanged joints are designed to remain in contact at accident temperature, concurrent with peak accident pressure.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 During refueling, the NPM, including the CNTS, is physically moved by a crane to the refueling area. The upper and lower CNV shells are separated during outages for refueling, maintenance, and inspection. The CNV is designed to accommodate comprehensive inspections of welds, including volumetric and surface inspections.

Welds are accessible, and there are no areas that cannot be inspected. The CNV design allows for visual inspection of the entire inner and outer surfaces. Through-wall degradation can be identified before development of potential leak paths precluding this as a credible leakage mechanism.

During reassembly, positive verification is used to verify proper stud elongation to ensure proper loading on each flange seal.

During normal operation, the CNV is partially submerged in borated water with its internal environment under near-vacuum conditions. Automatic engineered safety feature actuation systems initiate on high containment pressure. Containment vacuum pressure and leakage into the CNV is constantly monitored during normal operation. In comparison to traditional PWR designs, the small containment volume and evacuated operating conditions allows wide-ranging detection capabilities for liquid or vapor in-leakage, providing an additional layer of leakage integrity assurance.

The NuScale CNV design is described in detail in Section 3.0.

2.1 NuScale Containment Vessel Penetrations The CNTS design supports leakage integrity assurance through inspection and testing other than ILRT. When compared to traditional LLWR containments, the NuScale CNTS design is simplified. The CNV has a low number of penetrations, which are either ASME BPVC Class 1 flanged joints capable of Type B testing, ASME BPVC Class 1 welded nozzles with isolation valves capable of Type C testing, or part of a closed system inside containment (i.e., steam generator system piping). The CNV has no penetrations equipped with resilient seals. No instrument tubing penetrates containment; therefore, there are no small diameter fluid lines without isolation capability that are not subject to Type B or Type C LLRT. There are no air locks, flexible sleeves, or nonmetallic boundaries. The simplicity of design provides for alternate means of assuring containment leakage integrity. Leakage integrity assurance is primarily achieved by ensuring no unknown leak paths via ISI and accurately measuring the leakage rate of potential leak pathways via LLRT. Key features that ensure CNTS leakage integrity is maintained are:

Containment vessel flanges are designed to remain in contact at accident temperature, concurrent with peak accident pressure.

As described in Section 2.1, the CNV is an ASME Code Class 1 pressure vessel with a relatively low volume and no internal subcompartments. The NPM's comparatively simple design (compared to existing LLWR designs) allows identification of potential leakage pathways.

The CNV pressure vessel preservice test and inspections are equivalent to RPV requirements, including hydrostatic testing requirements. These tests and inspections verify no unknown leakage pathways exist.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Preservice design pressure leakage test of the CNV with CNV bolted flanges in place using as-designed flange covers installed with the design bolting materials, design bolting assembly preloads, and design seals to demonstrate no observed leakage at design pressure.

The upper and lower halves of the CNV are assembled for the first NPM of the initial NuScale plant.

After successful testing, the upper and lower halves of other CNVs may be tested separately.

Covers with electrical and instrumentation penetrations can be substituted with blank covers having the same sealing design.

The limited number of CNV penetrations have similar seal designs that are tested by Type B or Type C LLRT. The limited number of penetrations, and other aspects of the penetration design, allows accurate quantification of the overall leakage rate by LLRT.

The ISI Program and planned CNV examinations meet ASME Code Class 1 criteria.

This program ensures no new unidentified leakage pathways develop over time.

Disassembly and reassembly procedures and controls of the CNV are similar to the RPV. Positive verification is used to verify proper loading on each flange seal. This verification ensures these potential activity-based failure mechanisms do not degrade CNTS leakage integrity.

The CNV is an ASME BPVC Section III, Subsection NE, Class MC containment designed, fabricated, and stamped as a Subsection NB, Class 1 pressure vessel, with overpressure protection provided in accordance with NE-7000.

The CNV is made of corrosion-resistant materials, has a low number of penetrations, and features no penetrations with resilient seals. The use of welded nozzles and testable flange seals at the containment penetrations ensure Type B and Type C testing provide an accurate assessment of overall containment leakage rate.

The unique CNV and CNTS design allows testing and inspection options not suitable to current LLWR containment designs. Based on the containment vessel ASME pressure vessel design and its function, more alternate methods of testing and inspection are available. Each element of the CLIP is consistent with a corresponding element of an approved program for RPVs or LLWR containments.

The CNTS design is described in detail in Section 3.0. Table 2-1 compares elements of the CLIP with testing performed on the containment, reactor coolant pressure boundary (RCPB), and traditional containments. The table demonstrates the testing is commensurate with the design and safety function of the containment.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table 2-1 NuScale Containment Leak Rate Test Comparison Reactor Coolant CLIP Program Pressure Boundary Element to Ensure Traditional NuScale Containment Testing for NuScale and Essentially Containment Other Licensed Leak-Tight Barrier Facilities Initial verification of Hydrostatic testing per Hydrostatic testing per Preservice ILRT and structural integrity ASME BPVC Section III ASME BPVC Section III structural integrity test Initial verification of Factory - hydrostatic Hydrostatic testing per Preservice ILRT leakage integrity testing per ASME BPVC ASME BPVC Section III (leakage allowed below Section III Containment prescribed limit) preservice leakage test (ITAAC) (no visible leakage allowed)

On-site - preservice LLRT Prevention of leakage Administrative controls Administrative controls Administrative controls from activity-based such as configuration such as configuration such as configuration failure mechanisms management and management and management and (degradation due to procedural requirements procedural requirements procedural requirements system and/or for system restoration that for system restoration that for system restoration component ensure integrity is not ensure integrity is not that ensure integrity is modifications or degraded by plant degraded by plant not degraded by plant maintenance) modifications or modifications or modifications or maintenance activities maintenance activities maintenance activities Detection of leakage LLRT Reactor coolant system LLRT from activity-based (RCS) leak test -

failure mechanisms operational pressure Prevention of leakage Design and construction Design and construction Design and construction from age-based failure requirements for CNV, requirements for RCS, requirements, mechanisms inspections/ examinations inspections/examinations inspections/

(age-related performed in accordance performed in accordance examinations performed degradation) with ASME BPVC Section with ASME BPVC Section in accordance with XI, the maintenance rule XI, the maintenance rule ASME BPVC Section XI, and regulatory and regulatory the maintenance rule and commitments commitments regulatory commitments Detection of leakage NuScale CNV design RCS leakage detection ILRT from age-based failure allows for comprehensive mechanisms ISI surface and weld (age-related examination degradation)

Post-repair/ Hydrostatic testing per Hydrostatic testing per ILRT/LLRT modification ASME BPVC Section XI ASME BPVC Section XI verification of leakage and integrity LLRT Post-repair/ Hydrostatic testing per Hydrostatic testing per ILRT modification ASME BPVC Section XI ASME BPVC Section XI verification of structural integrity

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 2.2 Type B Testing Type B pneumatic tests detect and measure leakage across the pressure-retaining, leakage-limiting boundaries in the CNV. Preoperational and periodic Type B leakage rate testing is performed in accordance with 10 CFR 50, Appendix J, NEI 94-01 Rev. 3-A, and ANSI/ANS 56.8 within the test intervals defined by the licensee. The containment penetrations subject to Type B tests are identified in Appendix B. As described further in Section 3.2, the design of CNV penetrations allows accurate LLRT results to be obtained to quantify overall containment penetration leak rates.

The design of CNV Type B penetrations is described in Section 3.2.

2.3 Type C Testing The CIVs are designed to support Type C pneumatic tests. Preoperational and periodic Type C leakage rate testing of CIVs is performed in accordance with 10 CFR 50, Appendix J, NEI 94-01 Rev. 3-A, ANSI/ANS 56.8, and the licensee TS. The CIVs subject to Type C tests are identified in Appendix C. As described further in Section 3.3, the design of CIVs allows accurate LLRT results to be obtained to quantify overall CIV leak rates.

The design of CIVs is described in Section 3.3.

2.4 Containment Overall Leakage Limits Per 10 CFR 50, Appendix J, La is defined as the maximum allowable containment leakage rate in weight percent per day at peak containment accident pressure Pa. For the design, La is selected to be 0.20 weight percent of the containment air mass per day (over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) at the peak containment accident pressure (Pa) provided in FSAR Section 6.2.1. La is established as a safety analysis operational limit for the design. The values are used in calculations to confirm accident radiological containment leakage to the environment is within acceptable limits.

An evaluation of containment penetrations and access flanges (i.e., leakage pathways) was performed to determine whether the design can reliably meet the 10 CFR 50, Appendix J, leakage criteria using the maximum allowable leak rate of La of 0.2 weight percent of the containment air mass per day at design pressure. The evaluation concluded the combined maximum expected leakage from local penetrations, with conservative margin for degradation, is less than 0.60 La, at the peak accident pressure, Pa, which is the acceptance criterion for LLRT per 10 CFR 50, Appendix J.

Table 2-2 documents containment design basis leakage rate criteria. The CLRT leakage rate limits for LLRT are developed from these design basis limits to meet 10 CFR 50, Appendix J, leakage criteria.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Testing to meet less than 0.6La at Pa ensures the operational limit of 0.20 weight percent of containment air mass per day can be met. Design Specific Review Standard for NuScale Small Modular Reactor Section 6.2.1.1.A, Acceptance Criteria 4 states that to satisfy GDC 38 to rapidly reduce the containment pressure, the pressure should be reduced to less than 50 percent of the peak calculated pressure for the design basis loss-of-coolant accident within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the postulated accident. Following the peak containment pressure design basis accident, containment pressure drops from Pa to approximately 100 psia in less than five hours.

Table 2-2 Maximum Allowable Containment Leakage Rate Limits Leak Rate Pressure Containment leakage rate limit 0.2 weight percent of containment Pa air mass per day (La)

Containment leakage rate 0.2 weight percent of containment 1,200 psia (CNV internal design evaluation parameters air mass per day pressure)

The LLRT limits are developed based on the values of Table 2-2, and are based on a La at Pa and to meet less than 0.6La. The peak containment accident pressure (Pa) is identified in FSAR Section 6.2.1.The CLRT is described further in Section 7.0.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 3.0 NuScale Containment System Design The CNTS is designed around an ASME Code Class MC pressure vessel. The simplicity of the NPM design minimizes the number of containment penetrations required. There are a limited number of access ports (quantity nine), electrical penetration assemblies (EPAs, quantity 12), and instrument seal assemblies (ISAs, quantity four flanges with four modules per flange) that use the same flange seal design. The CNV closure flange, which separates the upper and lower CNV assemblies, uses the same seal design as the access ports. There are a limited number of fluid lines (quantity 14) penetrating containment (Figure 3-1). Of these, 12 are protected by CIVs and two are protected by a closed loop.

Figure 3-1 Containment Vessel Head

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 3.1 Containment Vessel Design The CNV is partially immersed in the ultimate heat sink (UHS) that removes residual core heat during normal and accident conditions. The design of the CNV features a relatively low and simple volume compared to other PWR containments: an approximately 6000 ft3 free volume steel vessel with no internal sub-compartments. The design prevents isolated pockets of concentrated gases. The upper and lower portions of the CNV are constructed of stainless steel. The CNV are shop-fabricated, which facilitates enhanced fabrication quality and testing control.

Figure 3-2 Containment Vessel

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 3.1.1 Preservice Tests and Inspections (Containment Vessel)

Nondestructive examination of the CNV after fabrication includes the following preservice examinations that are performed after hydrostatic testing but before code stamping:

general visual examinations for pressure retaining surfaces above the reactor pool level in accordance with Paragraph IWE-2200 VT-3 visual examinations for pressure retaining surfaces below the reactor pool level in accordance with Paragraph IWE-2200 VT-1 visual examinations for pressure retaining bolting in accordance with Paragraph IWE-2200 volumetric examinations for select welds in support of the break exclusion zone requirement in accordance with augmented requirements volumetric examinations for the: (a) CNV upper head to CNV upper seismic support shell, (b) CNV lower shell to CNV lower transition shell, and (c) CNV lower core shell to CNV lower head circumferential vessel welds in accordance with augmented requirements (note this inspection supports the Type A test exemption discussed in Section 1.3)

The CNV is hydrostatically tested after construction in accordance with ASME BPVC Section III, Paragraph NB-6000. The water-filled CNV is pressurized to a minimum of 25 percent over design pressure, and the pressure is held for at least ten minutes.

Pressure is then reduced to design pressure and then held for leakage examination.

The acceptance criterion for the test is that there are no leakage indications at the examination pressure (i.e., design pressure).

For the preservice design pressure leakage test, the CNV is tested with water at a minimum pressure of 1,200 psia to a maximum of 1,275 psia with the pressure held for 30 minutes prior to examining for leaks. The CNV bolted flange covers are attached to the vessel in their design condition during the preservice design pressure test. Covers with electrical and instrumentation penetration may be substituted with blank covers having the same sealing design as the design covers. The design seals are installed and the flanges are bolted using design bolting materials installed to normal operational preload values. Verification of bolting preload values is performed before vessel water-filled pressurization to design pressure. The testing is performed at a temperature between 70 degrees F and 140 degrees F which minimizes the possibility of brittle fracture. Once the vessel is at design pressure, the vessel bolted connections are visually examined for water leakage. The test is considered satisfactory if there is no indication of water leakage at the flange connections.

3.2 Preservice Tests and Inspections (Type B and Type C Components)

With the exception of the emergency core cooling system (ECCS) pilot valve body-to-bonnet seals (Type B penetrations), penetrations are either ASME Code Class 1 bolted flanged connections capable of Type B testing or ASME Code Class 1 welded nozzles with isolation valves capable of Type C testing.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 The CNV upper shell includes 12 EPAs and the CNV head includes four ISA flanges, as well as nine ports at various locations (Appendix B). The ISA flanges include four modules that are individually tested. The seal design for Type B penetrations, including the CNV closure flange is similar, with the exception being the size and model of the seals. These penetrations are tested periodically by Type B LLRTs.

Type C testing is required for eight CNV penetrations, seven of which are 2-inch nominal pipe size (NPS) pipe penetrations and one of which is 4-inch NPS (Appendix C). These CIVs are of identical design and construction. These penetrations are tested periodically by Type C LLRTs. The other six penetrations are main steam, feedwater, and decay heat removal system (DHRS) condensate penetrations that are connected to the steam generator (SG), which are closed loops inside containment. These penetrations are not required to be Type C tested in accordance with ANSI/ANS 56.8-2002, Section 3.3.1 and NEI 94-01 Rev. 3-A.

No instrument tubing penetrates containment; therefore, there are no small diameter fluid lines without isolation capability that are not subject to Type B or Type C LLRT. There are no air locks, flexible sleeves, or non-metallic boundaries. There are no penetrations in the NuScale design that would only be tested during an ILRT. Because of personnel safety and operational constraints, entry into the CNV does not occur during normal operation.

Type B pathway integrity is not expected to be disturbed except when the NPM is in a refueling outage or disassembled for emergent maintenance activities. Type C pathways are designed such that an individual valve can be tested in the same direction as the pressure applied when the valve would be required to perform its safety function. Type B and Type C pathways are tested to Pa.

NEI 94-01 Rev. 3-A provides guidance for implementing the performance-based option of 10 CFR 50, Appendix J (commonly referred to as Option B), for traditional large containment structures:

The purpose of Type A testing is to verify the leakage integrity of the containment structure. The primary performance objective of the Type A test is not to quantify an overall containment system leakage rate. The Type A testing methodology as described in ANSI/ANS-56.8-2002, and the modified testing frequencies recommended by this guideline, serves to ensure continued leakage integrity of the containment structure. Type B and Type C testing assures that individual leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. A review of performance history has concluded that almost all containment leakage is identified by local leakage rate testing.

Existing PWRs incorporate large containment building structures, however, leakage integrity is ensured in the NPM design by the welded-metal vessel design of the CNTS.

The CNTS is a small, high-pressure, ASME BPVC Class MC vessel with a significantly reduced number of penetrations and no internal sub-compartments. Preservice tests and inspections are similar to those performed on the RPV. Comprehensive ISI against ASME BPVC acceptance criteria also ensures continued system integrity. The surface areas and welds are accessible for inspection. Penetration pathways are tested to Type B or

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Type C criteria at accident or design pressures. This ensures leakage integrity of the CNTS is maintained without the need for Type A testing.

3.2.1 Type B Penetrations Type B components in the scope of the CLRT program are listed in Attachment A.1.

The CNV is designed to support Type B local leak rate tests to detect and measure leakage across pressure-retaining, leakage-limiting boundaries. The following containment penetrations are subject to preoperational and periodic Type B leakage rate tests.

flanged openings with bolted connections (i.e., access ports), nine total per NPM electrical penetration assemblies for various instrumentation and power cables, 12 total per NPM instrument seal assemblies for incore instrumentation, four total per NPM (each ISA includes four modules, resulting in a total of 20 Type B components)

ECCS trip/reset valve body-to-bonnet seals, four total per NPM containment vessel closure flange, one total per NPM Type B penetrations are bolted closures that have dual metal seals with testing ports between the seals. Type B penetration assemblies are designed and constructed to ASME Code Class 1 requirements. The CNV closure flange has a similar double seal and test port arrangement.

In addition to the penetrations listed above, most CIVs incorporate a test fixture with a removable cover that is subject to Type B testing. Section 3.3 provides details.

3.2.2 Access Ports The nine access ports are bolted closures that have dual metal seals with testing ports between the seals. Figure 3-3 shows the CNV head and control rod drive mechanism (CRDM) access flange. This double seal and test port design is used for every flange seal subject to Type B testing.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Figure 3-3 Containment Vessel Head / Control Rod Drive Mechanism Access Flange 3.2.3 Electrical Penetration Assemblies A total of 12 EPA penetrations are located at various locations on the CNV upper shell (Appendix B). The EPAs use sheath modules that use a glass-to-metal sealing technology that is not vulnerable to thermal or radiation aging, does not require periodic maintenance, and can achieve a less than minimum detectable leak rate.

The performance of the glass-to-metal EPA seal has been proven in currently operating nuclear plants.

The EPA, with installed modules, is bolted to CNV flange penetrations similarly to the flanged access ports. Figure 3-4 depicts the pressurizer (PZR) heater power supply EPA. This configuration is typical for EPAs in the design. The design includes the ability to test the double seals by pressurizing between the seals of the EPA similar to the flanged access ports. Sheath modules are only disassembled from an EPA for modification or if leakage is indicated. If disassembly is performed, then retest of the module or EPA seal is required before installing the EPA in the CNV.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Figure 3-4 Electrical Penetration Assembly Modules (Typical) 3.2.4 Instrument Seal Assemblies There are four ISA flanges located on the CNV head located at the incore instrumentation (ICI) penetrations. Each ISA flange contains four modules to accommodate ICI stringer assemblies and PZR level sensors. The ISA flanges and modules are leak tested separately.

The ISA contain ICI containment vessel compression seal fittings on its outer diameter for the 12 ICI stringer assemblies installed within the NPM, permitting the ICI stringer assemblies to penetrate the top head of the CNV while maintaining the containment pressure boundary during normal operations and design-basis events.

The ICI containment vessel compression seal fitting features a test port to allow local pneumatic pressurization of the region between the primary seal and test seals for Type B testing. The test port and test seals perform no function during normal plant operation.

3.2.5 Emergency Core Cooling System Trip and Reset Body-to-Bonnet Seals Four penetrations on the upper CNV shell accommodate ECCS valve trip/reset pilot assemblies. The trip/reset pilot valve body is located outside the CNV and is an RCPB. The ECCS valve trip/reset pilot assembly safe-end penetrations are welded to the external side of the penetration nozzle. Figure 3-5 shows the boundaries.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 The safe-ends and penetration nozzle-to-safe-end welds are part of the CNV. The valve assembly is welded to the penetration nozzle safe-end. The boundary is in the valve assembly-to-safe-end welds and the welds are part of the CNV. Pilot valve body-to-bonnet interfaces have a double seal with monitoring capability to allow Type B leakage testing. Trip and reset valve pressure boundaries are designed and constructed to ASME Code Class 1 requirements.

The ECCS trip valve and reset valve body-to-bonnet joint seals are not considered to be a flanged connection and are not included in the preservice design pressure leakage test or containment flange bolting calculation.

Figure 3-5 Emergency Core Cooling System Valve Trip/Reset Pilot Assembly (Simplified Diagram) 3.2.6 Containment Vessel Closure Flange The CNV closure flange allows disassembly of the CNV every outage for refueling, maintenance, testing, and inspection of the NPM. Figure 3-6 shows the CNV closure flange has a double seal with a test port. The design is similar to the flanged access ports. The test port attaches to tubing that runs from the CNV closure flange to an accessible point near the CNV manway access ports.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Figure 3-6 Containment Vessel Closure Flange Seal and Test Port 3.3 Type C Penetrations Type C components in the scope of the CLRT Program are listed in Appendix C.

Figure 3-7 shows a general depiction of these components, which are summarized below:

Primary system piping penetrations:

- CNV5: Reactor component cooling water system (RCCWS) supply to CRDMs

- CNV6: Chemical and volume control system (CVCS) injection

- CNV7: CVCS PZR spray

- CNV10: Containment evacuation system (CES)

- CNV11: Containment flooding and drain system

- CNV12: RCCWS return from CRDMs

- CNV13: CVCS discharge

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0

- CNV14: CVCS reactor coolant system high point degasification Secondary system piping penetrations:

- CNV1 and CNV2: Feedwater

- CNV3 and CNV4: Main steam

- CNV22 and CNV23: DHRS condensate return There are eight primary system piping penetrations into the CNV. Four piping penetrations (CNV6, CNV7, CNV13, and CNV14) are part of the RCPB and are subject to the requirements of GDC 55. These penetrations are protected by dual ASME BPVC Class 1 primary system containment isolation valves (PSCIVs) of identical design. The other four piping penetrations (CNV5, CNV10, CNV11, and CNV12) are open to containment atmosphere and are subject to the requirements of GDC 56. The RCCWS is conservatively considered to be open to containment. These ASME BPVC Class 2 penetrations are protected by dual ASME BPVC Class 1 PSCIVs of identical design, with the exception of the CES penetration, which has ASME BPVC Class 2 valves.

There are six secondary system piping penetrations into the CNV, none of which penetrations require Type C testing. These six penetrations are open to a closed loop inside containment (the SGS) and are subject to the requirements of GDC 57. Four of these penetrations (CNV1, CNV2, CNV3, and CNV4) are protected by single ASME Code Class 2 secondary system containment isolation valves (SSCIVs) and nonsafety-related backup valves. The other two penetrations (CNV22 and CNV23) are protected by an ASME Code Class 1 and Class 2 closed loop inside containment and an ASME Code Class 2 closed loop outside containment. The DHRS penetrations do not feature CIVs.

The design supports an exemption from GDC 57 to clarify the use of a closed loop system inside and outside containment.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Figure 3-7 Primary System Containment Isolation Valve and Secondary System Containment Isolation Valve Arrangement on Containment Vessel Head

((2(a),(c) The CIVs on both primary and secondary systems are quarter-turn ball valves with the same actuator design. The size and ball design varies between primary and secondary valves.More differences are found in the main steam isolation valves (MSIVs) and feedwater isolation valves (FWIVs), but the majority of design features are identical. The CIV pairs feature a containment isolation test fixture (CITF) valve located between the shared valve body and the CNV head (Figure 3-8.), with the exception of the MSIVs and associated bypass valves. The CITF valve is a single, top-entry ball valve that features a test port to accommodate Type C LLRTs of the downstream CIVs. The CITF valve features a cover that is keyed to the ball such that the CITF valve remains locked open during operation and the cover must be removed and rotated to close the CITF valve for testing. This cover features a double o-ring cap-to-body seal with a test port for Type B pneumatic leakage testing of the seal. The leak rate measured on each CITF © Copyright 2022 by NuScale Power, LLC 27

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 valve cover following maintenance and testing is included in the combined Type B and Type C leak rate summation. 3.3.1 Primary System Containment Isolation Valves The PSCIVs are the only piping penetration isolation valves required to meet 10 CFR 50, Appendix J, Type C test criteria. Both GDC 55 and GDC 56 penetrations are protected by PSCIVs of identical design and construction. These PSCIVs are 2-inch NPS (except for two valves on the containment evacuation system, which are 4-inch NPS) and have a dual-actuator, single-body arrangement. Four PSCIVs protect GDC 55 penetrations, and four PSCIVs protect GDC 56 penetrations. The PSCIVs are designed and constructed to ASME Code Class 1 except the two CES PSCIVs, which are ASME Code Class 2 valves. Figure 3-8 Primary System Containment Isolation Valve Dual Valve, Single Body Design © Copyright 2022 by NuScale Power, LLC 28

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Each dual valve assembly is welded directly to its respective CITF, which in turn is welded to the CNV head via integral vessel nozzle safe end and butt weld arrangements. The containment flooding and drain system and RCCWS valves are classified as Quality Group B, but are specified to be designed and constructed to ASME BPVC Class 1. 3.3.2 Secondary System Containment Isolation Valves 3.3.2.1 Feedwater Isolation Valves There are two feedwater containment penetrations. Both of these piping penetrations are 5-inch NPS. Each penetration is protected by a dual ASME BPVC Class 2 SSCIV. The FWIV has a dual valve, single body arrangement. Each assembly consists of an actuated isolation valve and an integral self-actuating check valve. The dual valve assembly is welded to a CITF that is welded directly to the CNV head via integral vessel nozzle safe end and butt weld arrangements. The isolation valve is a hydraulic-to-open, ball valve that uses a stored energy device to close. It is the inboard valve in the dual valve arrangement. The outboard valve is a safety-related nozzle check valve. The function of the check valve is to close more rapidly than the inboard isolation valve to preserve DHRS inventory in the event of a feedwater line break outside containment. The FWIV has no containment isolation function and no other specific leakage criteria. Neither the FWIV nor the feedwater isolation check valve has a Type C test requirement as defined by the IST Plan, which is typical for FWIV of a PWR design; however, the FWIV is classified as IST Category A because of the specific leakage criteria needed for DHRS operability. As such, the CITF on each FWIV is used for IST purposes and no Type B test is performed on the cover. © Copyright 2022 by NuScale Power, LLC 29

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Figure 3-9 Feedwater Isolation Valve Dual Valve, Single Body Design 3.3.3 Steam Generator Thermal Relief Valves Two thermal relief valves are provided inside containment with one on each of the SG closed loops. These relief valves provide secondary side overpressure protection during chemistry control evolutions in support of startup and shutdown. Feedwater lines, SG, and main steam lines could experience water solid conditions if the SSCIVs inadvertently close during these evolutions with decay heat in the reactor core. The thermal relief valves are designed to relieve thermal overpressure during water solid conditions to maintain the integrity of the SG closed loops. The valves are installed on flanges to facilitate removal for periodic replacement as established by the IST Program. These valves form part of the SG closed loop boundary as part of the containment boundary. However, they are not leak tested per 10 CFR 50, Appendix J, because these valves are ASME BPVC Class 2, Quality Group B. relieve into the CNV. have a set pressure approximately 1,000 psia above peak operating pressure (2,200 psia vs. 1,200 psia). are leak tested pursuant to ASME Operation and Maintenance Code Mandatory Appendix I with the licensee leakage criteria for replacement valves specified so that reliable operation can be reasonably established throughout the ten-year test interval. © Copyright 2022 by NuScale Power, LLC 30

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 3.3.4 Main Steam Isolation Valves and Bypass Valves There are two main steam containment penetrations, both of which are 12-inch NPS. Each penetration is protected by an ASME BPVC Class 2 SSCIV with an integral bypass valve on a 2-inch NPS bypass line. Both the MSIVs and the main steam isolation bypass valve (MSIBV) have a single actuator, single body arrangement. The MSIBV is integral to the MSIV in a parallel arrangement but operated independently. The MSIV and MSIBV are hydraulic-to-open,ball valves that use a stored energy device to close. Each valve assembly is welded directly to the CNV head via integral vessel nozzle safe end and butt weld arrangements. The functions of the MSIV and the MSIBV are containment isolation, main steam isolation, and DHRS boundary during DHRS actuation. Neither the MSIV nor the MSIBV has a 10 CFR 50, Appendix J, Type C test requirement as defined by the IST Plan. However, both the MSIV and the MSIBV are classified as IST Category A because of the specific leakage criteria needed for DHRS operability. Thus, the leak testing features on these valves is only used for IST purposes. Figure 3-10 CombinedMain Steam Isolation Valve and Main Steam Isolation Bypass Valve, Single Unit Design 3.3.5 Feedwater Plenum Cover Drain Valves Each SG feedwater plenum cover is equipped with a drain valve to facilitate draining the SG during outage maintenance evolutions. As they are used only for system maintenance, they are exempt from the IST Plan, as well as 10 CFR 50, Appendix J testing. © Copyright 2022 by NuScale Power, LLC 31

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 4.0 Inservice Inspection and Testing of Containment Inservice inspection and IST are required by 10 CFR 50.55a(g) and (f), respectively, to ensure periodic requisite inspection and testing is performed on the CNTS that ensures leak-tight integrity is maintained. Type B testing is specified in the ISI Plan and Type C testing in the IST Plan. Both the ISI and IST Programs are an integral part of the CLRT Program. 4.1 Inservice Inspection Inservice inspection provides an essential function for the CLRT Program by confirming CNTS integrity and ensuring no new leakage paths are present as discussed in the following sections. 4.1.1 Weld Inspections The specified surface, volumetric (ultrasonic), and visual examinations within the ISI Program provide reasonable assurance that no new leakage paths will develop over the service life of the NPM. The CNV design allows comprehensive inspections of welds, including volumetric and surface inspections. The CIVs are located outside the CNV and pressure boundary welds are accessible with no areas that cannot be inspected. The reduced ISI requirements from ASME BPVC Section XI for small primary system pipe welds between the CNV and the CIV are not applied to these welds. Welds between the CNV and the CIV are inspected each test interval. 4.1.2 Visual Inspections ASME BPVC, Class MC, Section IWE requires visual examination for structures, systems, and components subject to normal degradation and aging. Surface areas subject to accelerated degradation and aging require an ultrasonic thickness exam. Additionally, based on the high pressure and safety functions of the CNV, the ISI Program requires augmented examinations of the CNV in accordance with ASME Code, Class 1 requirements. The CNV design allows visual inspection of the entire inner and outer surfaces; therefore, developing an undetected leak through the metal pressure boundary is unlikely. 4.1.3 Bolting Inspections Inspection of CNV bolting is required per ASME BPVC, Section XI. For bolting provided for the CNV closure flange (i.e., CNV main closure studs), these inspections are limited to visual examination (VT-1) per ASME BPVC, Section XI, Subsection IWE-2500, Category E-G for metal containments. Bolting of other flanges installed on the CNV are subject to visual examination (VT-1) per ASME BPVC, Section XI, Subsection IWE-2500, Category E-G. These examinations may be performed without disassembling the joint; however, if the joint © Copyright 2022 by NuScale Power, LLC 32

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 is disassembled during the inspection interval, the examination is required to be performed with the joint disassembled. 4.1.4 Steam Generator Inspections and Controls Each SG forms part of a GDC 57 closed loop containment barrier for PWRs; therefore, the reliability of its integrity and its failure mechanisms contribute to the reliability of the containment boundary. The NuScale SG design is different from traditional SGs. Major differences include: The SG is located inside the RPV and is not a separate component attached by RCS piping. The tubes are helically coiled in the annular space between the walls of the RPV and a concentric upper riser internal to the RPV. Steam is generated on the inside of the tubes with lower steam pressure inside the tube and higher RCS pressure on the outside. As a GDC 57 closed loop system, each SG is isolated by single SSCIV (FWIV on the inlet, MSIV and MSIBV on the outlet). The SG is an ASME BPVC Class 1 RCPB. Detailed inspection requirements for the SG tubing are part of the ISI program. Technical Specification Section 5.5.4 establishes a Steam Generator Program to ensure SG tube integrity is maintained. 4.1.5 Type B Testing Type B testing is local pneumatic pressure leak rate testing of containment penetrations, specifically the EPAs, IPAs, access ports, ECCS pilot valve bodies, and the CNV closure flange. These tests are inservice inspections specified in the ISI Plan. The ISI Program specifies Type B LLRTs. 4.2 Inservice Testing The IST Plan identifies valves in the scope of the IST Program with specific leakage criteria. Valves with specific leakage criteria as a containment boundary are identified as LTJ, which denotes a valve with a 10 CFR 50, Appendix J, Type C leakage test requirement. The IST Plan also specifies test frequencies pursuant to the requirements of NEI 94-01 Rev. 3-A. © Copyright 2022 by NuScale Power, LLC 33

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 5.0 Type B Local Leak Rate Testing Type B tests of the double seals on the containment bolted closures are performed by local pressurization at containment peak accident pressure, Pa. Pressurized gas such as air or nitrogen is applied to the test ports, which are provided between the two seals in each bolted closure and the pressure decay over time or the leak flow rate is measured. Type B tests use either the pressure decay or flow makeup method of detection. For the pressure-decay method, a test volume is pressurized with air or nitrogen to at least Pa. The rate of decay of pressure in the known test volume is monitored over time to calculate a leakage rate using the pressure-decay method. For the flow makeup method, the required test pressure is maintained constant in the test volume by making up test fluid, such as air or nitrogen, through a calibrated flowmeter. With pressure held constant, the makeup flow rate is equivalent to the leakage rate from the test volume. The design combined leakage rate for penetrations and valves subject to Type B and Type C tests is limited to less than 0.6La. An overall leak rate of less than 0.6La (Section 7.2) is confirmed by LLRT before the startup of each NPM. In accordance with 10 CFR 50, Appendix J, Type B tests are performed during each reactor shutdown for refueling or at other convenient intervals in accordance with the CLRT program. 5.1 Type B Test Method Type B penetrations are tested each refueling outage. An as-found test is required to be performed before any Type B penetration is opened or manipulated in a way that would affect the leak-tightness of the penetration (Section 7.5 has a discussion of test considerations, including preconditioning). Test equipment is installed on the test port located between the double seals. The seal is then tested with compressed air or nitrogen using either the pressure decay or flow makeup method to measure the leakage as specified in the CLRT program. Once as-found testing is performed and documented, the penetration can be opened. Just inside the CNV head manway is a small tubing connection to the CNV closure flange test port. The Type B test rig is connected at this point and an as-found test of the CNV closure flange is performed (Figure 3-6). Once the refueling outage is completed and penetrations are closed for the final time, an as-left Type B test is performed on penetrations. If a penetration was not opened and no bolts were manipulated, and the as-found test was within CLRT acceptance criteria, then the as-found test may be credited as the as-left test with no further testing needed. The CNV closure flange is tested twice after it is reassembled. The first as-left test occurs in the refueling area to ensure the new CNV closure flange seals are installed properly and are sealed. The second (final) as-left test occurs after the NPM is moved to the operating bay. The second as-left test ensures CNV movement had no adverse effect on the leak-tightness of the CNV closure flange seal. After the CNV closure flange seal is tested in the operating bay, then the CNV head manway cover is reinstalled and tested. Section 7.5 contains additional discussions on as-found and as-left testing. © Copyright 2022 by NuScale Power, LLC 34

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 5.2 Electrical Penetration Assemblies The EPA sheath modules are installed and tested at the shop. Glass-to-metal seals (penetrations), exclusive of the flange-to-nozzle seals, are designed for leakage rates not to exceed 1.0 x 10-3 standard cm3/s [1.27 x 10-4 standard cubic foot per hour (scfh)] of dry nitrogen at design pressure and ambient temperature. Glass-to-metal seals typically achieve leak rates in the undetectable range, or 1.0 x 10-7 standard cm3/s of dry nitrogen at design pressure and ambient temperature. The glass-to-metal module seal is an established sealing technology that is not vulnerable to thermal or radiation aging and does not require periodic maintenance or testing. This module-to-EPA seal does not require periodic testing. It would only be tested after completing maintenance activities that affect the seal. The EPA flange seal is the same double o-ring seal design of all Type B penetration seals. The required installation acceptance criterion for leakage rate of each EPA penetration is 1.0 x 10-2 standard cm3/s (1.27 x 10-3 scfh) per Institute of Electrical and Electronics Engineers Standard 317-1983 (Reference 9.5). The leakage margin allotment for Type B testing is preliminarily selected to be (( }}2(a),(c) the installation acceptance criterion. With this allotment, the EPA contribution to the overall containment leakage rate does not challenge the acceptance criterion of 0.6La. 5.3 Instrument Seal Assemblies The ICI containment vessel compression seal fitting is a mechanical seal device that contains a primary seal, which is solely responsible for maintaining the CNV pressure boundary at the outside diameter of the ICI stringer assembly. The primary seal uses metal ferrules to form a leak tight compression seal with the OD of the ICI stringer assembly. Mechanical requirements are specified to aid with proper swaging of the ferrules, ensuring a leak tight seal is formed when properly installed. A leakage test port in the ICI containment vessel compression seal fitting allows local pneumatic pressurization of the region between the primary seal and test seals to allow Type B testing. The test port and the test seals perform no function during normal plant operation. 5.4 Access Ports The CNV access port flange seals feature an identical double seal design. The leakage performance of these seals is similar to the EPA flanges based on an evaluation of leakage performance for off-the-shelf metal seals. 5.5 Emergency Core Cooling System Pilot Valve Bodies There are four 3-inch NPS containment penetrations for the ECCS trip and reset valve assemblies (two of each). A Type B test is required at the double seal between the valve bonnet and body (Figure 3-5). The rest of these valve bodies are self-contained metal barriers that form part of the containment pressure boundary. Leakage criteria for these seals are lower than other Type B boundaries because of the smaller size of the seals. © Copyright 2022 by NuScale Power, LLC 35

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 5.6 Containment Vessel Closure Flange The CNV closure flange is a large, double seal design with an approximately 45-foot circumference. This seal maintains the containment boundary between the upper and lower CNV assemblies. The CNV closure flange leakage limit for the CLRT program is estimated to be (( }}2(a),(c) based on the linear seal length and performance of off-the-shelf metal seals. 5.7 2Bolting The CNV bolting design for the closure flange and EPAs, instrument seal assemblies, and access ports is in accordance with ASME BPVC, Section III, Division 1, Subsection NB. The preload needed to maintain a tight joint maintains seal integrity with design pressure in containment. Preload requirements are: The bolt preload for the design pressure is sufficient to resist the hydrostatic end force and maintain a compression load on the gasket contact surface to ensure a tight joint when the design pressure is applied to the internal surfaces. Preload is applied to the joint at atmospheric conditions without the presence of internal pressure. The CNV bolted closure design and preload design requirements ensure Type B flange seals, including EPAs and ISAs, remain in contact at accident temperature concurrent with peak accident pressure. Flanges are as-found tested in accordance with 10 CFR, 50 Appendix J, before removal for refueling outage activities. The licensee's administrative controls are used during reassembly, including preload verification and quality control hold points, to ensure EPAs, ISAs, access ports, and flange seals are reassembled with fasteners at the correct preload. An as-left Type B test on the penetration seal verifies leakage is within the CLRT program limit. © Copyright 2022 by NuScale Power, LLC 36

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 6.0 Type C Local Leak Rate Testing The PSCIVs are tested using either the pressure decay or flow makeup method at containment peak accident pressure, Pa. For the pressure decay method, the test volume is pressurized with air or nitrogen. These test methods are described in Section 5.0. Pressure to the PSCIV is applied in the same direction as the pressure applied when the valve is required to perform its safety function. The CITF valve is closed to provide a test boundary upstream of the PSCIV. There are no Type C leak test requirements for the SSCIVs (FWIVs, MSIVs, and MSIBVs), although these valves do have specific leakage criteria for DHRS operability. Leak testing of the SSCIVs is in accordance with the TS and the IST Plan to maintain DHRS operability. The leak testing features on these valves is used only for these purposes. 6.1 Type C Test Method Each CIV to be tested is closed by normal means without preliminary exercising or adjustments (Section 7.5 has a discussion of test considerations, including preconditioning). This closure can be achieved via the periodic closed-stroke test required by the IST Program. Piping is drained and vented as needed and a test volume is established to produce a differential pressure across the valve when pressurized. The CIV is then prepared for testing by removing the cover of the upstream CITF valve and closing the internal ball valve, then reinstalling the cover to lock the ball valve in place. The test port on the CITF assembly is then used to establish test pressure in the same direction as the pressure applied when the CIV would be required to perform its safety function (i.e., upstream from the CNV). Test equipment is installed on the test port and system valves are aligned so a vent path is established downstream of the tested valve. The CIV is then tested via air or nitrogen using either the pressure decay or flow makeup method as specified in the CLRT program. When testing the first CIV in the penetration is completed, the test equipment is vented and the valves are realigned. The first CIV is opened and the second CIV is closed to establish the test alignment for the latter.The test equipment is re-pressurized and the second CIV is then tested. Once the LLRT is completed on the inboard and outboard CIVs, the system is vented and the test equipment is disconnected from the CITF test port. Then, the desired post-test system lineup is established, including opening the CITF ball valve and re-orienting the cover to lock it in place. © Copyright 2022 by NuScale Power, LLC 37

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 7.0 Containment Leakage Rate Test Program The CLRT program contains the following attributes: Limits are applied that are established in the plants design basis and the TS to establish LLRT criteria to ensure penetrations meet the preservice and periodic limit of 0.6La at Pa for the combined leakage rate of penetrations and valves subject to Type B and C tests. Type B LLRT is performed in accordance with the ISI Plan frequency. Type C LLRT is performed in accordance with the IST Plan frequency. The results of containment system ISI are documented. The results of found and as-left Type B and Type C LLRTs are documented. Post-maintenance testing results on a Type B and Type C pressure boundary are documented. An adverse condition is analyzed for generic considerations. Type B seals are the same double seal design, and Type C valves are the identical 2-inch design, except the two containment evacuation system CIVs, which are 4-inch. Additionally, each site employing the design has six identical NPMs and CNTS. Records are maintained to produce periodic leakage test summary reports that are available onsite for NRC review in accordance with NEI 94-01 Rev. 3-A. 7.1 Challenges Associated with Type A Testing Besides the GDC 52 exemption basis discussed above, the actual performance of Type A integrated leak rate testing on an NPM poses significant challenges that render the test either invalid or infeasible. These challenges include but are not limited to temperature variations due to continuous heat transfer between the CNV (which is normally filled with water when fully assembled in the operating bay during refueling outages) and the UHS in which it is mostly immersed, as well as from core heat from the fuel at the bottom of the RPV. the procurement, arrangement, and installation of a multitude of high-precision temperature sensors beyond normally installed instrumentation to ensure temperature variations are properly detected and compensated in both liquid and air spaces. greater accuracy requirements for test instrumentation given the larger proportional impact of temperature changes and instrument errors on pressure at the magnitude of Pa, which is much higher compared to traditional PWR designs. the lack of available, off-the-shelf sensors capable of measuring dew point temperature and relative humidity at high-pressure, no-flow conditions that can also be arranged and calibrated per the requirements of Reference 9.5. the application of standard ILRT acceptance criteria (75 percent of La) to a much smaller CNV volume, resulting in an exceedingly low allowable leak rate that may not © Copyright 2022 by NuScale Power, LLC 38

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 be achievable during the ILRT, especially considering the increased effects of instrument errors. Appendix D provides additional discussion on these challenges. 7.2 Containment Leakage Limits The leak rates of penetrations and valves subject to Type B and Type C testing are combined in accordance with 10 CFR 50, Appendix J. The combined leakage rate for penetrations and valves subject to Type B and Type C tests shall be less than 0.6La at Pa. The CLRT limits are derived from the design basis limits to meet 0.6La for LLRT. If repairs are required to meet CLRT limits, the results are reported in a separate summary to the NRC, in accordance with 10 CFR 50, Appendix J, to include the structural conditions of the components that contributed to the failure. As each Type B or Type C test (or group of tests) is completed, the combined total leak rate is revised to reflect the latest results. Thus, a reliable summary of containment leak-tightness is maintained current. Leak rate limits and the criteria for the combined leakage results are described in the plant TS. 7.3 Test Frequency Schedules for performing periodic Type B tests are specified in the owners ISI Plan and periodic Type C tests are specified in the owners IST Plan. The CLRT Program is endorsed in the plant TS Section 5.5.9. Provisions for reporting test results are described in the CLRT Program. Conditional testing is in accordance with the owners procedures, but includes Type B or Type C testing when repair, replacement, or modification to a containment pressure boundary takes place. Upon initial startup of each NPM, Type B tests are performed during reactor shutdown or refueling, or at other convenient intervals, but in no case at intervals greater than two years (as specified in the owners ISI Plan) per 10 CFR 50, Appendix J, Option B. Type C tests are performed during reactor shutdown or refueling, but in no case at intervals greater than 30 months (as specified in the IST Plan) per 10 CFR 50, Appendix J, Option B and NEI 94-01 Rev 3-A. Performance-based (i.e., extended) test frequencies under 10 CFR 50, Appendix J, Option B can be adopted for Type B and Type C tests once satisfactory performance is established. Section 7.5 provides further details. 7.4 Test Results and Reporting Requirements The CLRT Program reporting requirements are pursuant to 10 CFR 50, Appendix J, Option B. Preoperational and periodic tests are documented in a summary report that is made available for inspection, upon request, at the plant site. The summary report includes, at a minimum, a schematic arrangement of the leakage rate measurement system, the instrumentation used, the supplemental test method, the test program © Copyright 2022 by NuScale Power, LLC 39

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 selected as applicable to the preoperational test, and subsequent periodic tests. The report contains an analysis and interpretation of the leakage rate test data for the Type B and Type C test results and the applicable ISI results to the extent necessary to demonstrate the acceptability of the containment leakage rate in meeting acceptance criteria. For each periodic test, leakage test results from Type B and Type C tests are included in the summary report. The summary report contains an analysis and interpretation of the Type B and Type C test results and the applicable ISI results that were performed since the last inspection interval (usually the last refueling outage). Leakage test results from Type B and Type C tests that failed to meet the CLRT Program acceptance criteria are included in a separate accompanying summary report that includes interpretation of the test data, and the structural conditions of the containment or components, if any, that contributed to the failure in meeting the acceptance criteria. If performance-based (i.e., extended) test frequencies under 10 CFR 50, Appendix J, Option B are adopted for Type B and Type C tests, additional reporting requirements are imposed by NEI 94-01, Rev. 3-A. To facilitate the transition from baseline frequency testing (30 months) to extended intervals, these requirements can be included as part of standard CLRT Program reporting from initial startup. Section 7.5 provides additional details. 7.5 Special Testing Considerations 7.5.1 As-Found Testing As-found testing is performed to determinehow a component would perform if called upon in an accident scenario. It is the first actuation of the component that has been in standby mode since the last time it was tested or for normal operation. Technical specification limiting conditions for operation criteria are generally as-found values. The as-found results are used to determine whether the leak-tightness of a component and the overall containment degraded over time. 7.5.2 As-Left Testing As-left testing is the final performance of a surveillance test or calibration of a component to determine its functional performance before placing it back into service. Technical specification surveillance test criteria are generally as-left values. 10 CFR 50, Appendix J, LLRT, requires as-found testing of Type B and Type C penetrations when entering a refueling outage, and as-left testing when reassembling Type B penetrations or performing post-maintenance testing on a PSCIV (if maintenance was performed that affected the leak tightness of the valve). The as-left results establish operational readiness until the next scheduled LLRT and ensure the leak-tightness of a component (and the overall containment) does not degrade to an unacceptable level over time. © Copyright 2022 by NuScale Power, LLC 40

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 7.5.3 Preconditioning Preconditioning occurs when a component is exercised, adjusted, or otherwise manipulated before as-found testing performance. The ASME and NRC requirements do not allow preconditioning for the performance of any as-found testing. For Type C penetrations, the owner balances the requirements of the IST Program and the CLRT Program. As-found stroke time testing is required for IST and as-found LLRT is required for 10 CFR 50, Appendix J. These tests are coordinated to preclude preconditioning. 7.5.4 Reverse Direction Testing Special considerations from References 9.4and 9.6 apply when testing a component in the reverse direction (i.e., applying test pressure in the opposite direction of post-accident pressure). For the design, no reverse direction testing is currently performed, but these requirements are reviewed if any reverse testing becomes necessary as the design matures. 7.5.5 Modifications After Preoperational Testing Any major modification or replacement of components that are part of the containment pressure boundary performed after preoperational leakage rate testing are followed by a Type B or Type C test as applicable for the area affected by the modification. The measured leakage from the test is included in the summary report. 7.5.6 Option B Performance-Based Testing Applicants that reference the NuScale power plant US460 standard design approval may not be able to initially adopt the test method frequencies specified in 10 CFR 50, Appendix J, Option B, Performance-Based Requirements. However, the licensee is expected to adopt Option B once sufficient operating history is obtained under Option A to use this performance-based approach. Multi-module testing does not impact test frequencies of the owners CLRT Program. Multi-module testing does also not affect the test frequencies of either the ISI or IST Programs. Risk-informed methods are not available to initial ISI or IST Programs, yet multi-module testing is a factor in CLRT, ISI, and IST. Generic considerations of adverse conditions not only potentially affect similar components in the affected NPM. Consideration must also be given to similar components across the NPMs. A plant with six NPMs nominally plans for three refueling outages annually. This outage frequency provides a rapid accumulation of performance history for the CLRT, ISI, and IST Programs. With NRC approval, risk-informed methods could be applied sooner compared to a traditional one- or two-reactor plant. Under Option B, the performance-based testing requirements established by NEI 94-01 Rev 3-A applies for Type B and Type C local leak rate testing. To summarize these requirements: © Copyright 2022 by NuScale Power, LLC 41

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 The LLRT intervals for Type B components are increased from the baseline frequency of 30 months up to a maximum of once per 120 months (10 years) following the completion of two consecutive periodic as-found tests that satisfy administrative limits. The LLRT intervals for Type C components are increased from the baseline frequency of 30 months up to a maximum of once per 75 months following the completion of two consecutive periodic as-found tests that satisfy administrative limits. Though not required, the expected industry best practice is to obtain three satisfactory as-found tests before adopting a 75-month extended LLRT interval. If a 75-month interval is adopted for Type C LLRTs, the Limitations and Conditions provided in Reference 9.6 apply. If a valve is replaced or engineering judgment determines that modification of a valve has invalidated the valve performance history (for example, replacement of a part that affects seat tightness), the valve is tested at the baseline frequency of 30 months. For both Type B and Type C components, as-found testing is performed before a maintenance, repair, modification, or adjustment activity if the activity could affect the penetrations leak tightness. An as-left test is performed following maintenance, repair, modification, or adjustment activities. If results are not acceptable, then the testing interval is set at the baseline frequency, and a cause determination is performed and corrective actions identified that focus on activities that can eliminate the identified cause of failure with appropriate steps to prevent recurrence. A post-outage report is prepared presenting results of the previous cycle Type B and Type C tests, as well as Type B and Type C tests performed during that outage. The technical contents of the report are generally described in Reference 9.4 and are available on-site for NRC review. The report documents that the applicable performance criteria are met, and serves as a record that continued performance is acceptable. The report includes the combined Type B and Type C leakage summation, and the margin between the Type B and Type C leakage rate summation and its regulatory limit (Topical Condition 1 discusses further). Adverse trends in the Type B and Type C leakage rate summation are identified in the report and a corrective action plan is developed to restore the margin to an acceptable level. Certain limitations apply as described in Section 11.0 of Reference 9.6. Beyond this standard, the following considerations also inform the owners Option B testing plan: An as-found test is performed before work is done that can affect the leak rate of a component whose leakage integrity is suspect (Section 3.3.4.1 of Reference 9.4). © Copyright 2022 by NuScale Power, LLC 42

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 In the event mid-cycle maintenance is required between refueling outages for a component on an extended LLRT interval, the as-left test from the last refueling outage may be counted as the as-found test. Extended LLRT intervals need not be short-cycled to accommodate Type A tests (i.e., ILRT) because an exception is being pursued for these tests as discussed in Section 2.0. Additionally, the NRC Safety Evaluation Report incorporated into NEI 94-01 Rev. 3-A imposes two additional conditions that must be considered for Option B testing. Topical Condition 1 requires if a 75-month extended interval is adopted for Type C tests, then (1) a licensees post-outage report shall include the margin between the Type B and Type C leakage rate summation and its regulatory limit and (2) a corrective action plan shall be developed to restore the margin to an acceptable level. Topical Condition 2 requires if a Type C test interval is extended beyond 60 months (i.e., up to 75 months), then CLRT program trending and monitoring must include an estimate of the amount of understatement in the Type B and Type C leakage rate summation that must be included in a licensees post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations. The common industry practice is to include an additional 25 percent of the measured leak rate on any Type C component on a 75-month interval in the Type B and Type C leakage rate summation (e.g., if a particular valve has a measured leak rate of 0.1 scfh, a value of 0.125 scfh would be included in the total). Upfront consideration of Option B requirements can facilitate the eventual transition from baseline frequency testing (30 months) to extended intervals if CLRT Program documentation, procedures, and tools used for scheduling, trending, and monitoring are initially developed with these requirements already (or readily able to be) incorporated. © Copyright 2022 by NuScale Power, LLC 43

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 8.0 Material Selection and Aging Degradation Leakage Rate Test Program The containment design includes material selection that supports leakage integrity assurance. Potential degradation can be identified by inspection and examination before the formation of containment leakage pathways. 8.1 Material Selection and Operating Conditions Table 8-1 lists the CNV pressure-retaining materials. Nonpressure-retaining materials are not included or discussed because they are not involved in maintaining the leak tightness for the CNV. Most of the CNV is immersed in reactor pool water during plant operation. The CNV operating temperature is 100 degrees F. Although the minimum specified pool water temperature is 65 degrees F, the typical pool water temperature is approximately 100 degrees F under operating conditions. During the plant shutdown process, the CNV is flooded with reactor pool water when the operating condition is in the safe shutdown mode and the RCS coolant temperature drops below 300 degrees F. The portion of the CNV in contact with RCS coolant during plant operation is the only portion that is part of the RCPB. The CNV components that are in contact with RCS coolant include the following nozzles and their safe ends: CVCS injection nozzle (CNV6) CVCS discharge nozzle (CNV13) PZR spray nozzle (CNV7) RPV high point degasification nozzle (CNV14) The CNV (shells, flanges, top head, nozzles, and covers) is made of Grade F6NM martensitic stainless steel that is included as an allowable material by Regulatory Guide (RG) 1.84 by approval of ASME BPVC Case N-774 or SA-182 except for a portion of the lower CNV. The lower CNV (lower head, core region shell, and transitional shell) with peak 60-year design fluence exceeding 1E+17 n/cm2, > 1 million electron volts (MeV) is made of SA-965 Grade FXM-19 austenitic stainless steel. © Copyright 2022 by NuScale Power, LLC 44

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table 8-1 Containment Vessel Pressure-Retaining Materials Item Material CNV Vessel CNV top head cover SA-182 Grade F6NM CNV top head and upper CNV shells and flange ASME BPVC Case N-774 (SA-336) Grade F6NM Lower CNV (flange and lower shell) ASME BPVC Case N-774 (SA-336) Grade F6NM Lower CNV (lower head, core region shell, and transition shell) SA-965 Grade FXM-19 (Note 2) Nozzles and Access Ports Nozzles and access ports SA-182 Grade F6NM Safe-ends for nozzles SA-182 Grade F304 (Note 1) Covers for access ports SA-182 Grade F6NM Pressure-Retaining Bolting Bolting for CNV main closure flange SB-637 UNS N07718 (Note 3) Bolting for other than CNV main closure flange SB-637 UNS N07718 (Note 3) SA-564 Type 630 Condition H1100 SA-193 Grade B8 Class1 SA-194 Grade 8 Weld Filler Metals 2XX austenitic stainless steel weld filler metals SFA-5.4: E209, E240 (Note 2) SFA-5.9: ER209, ER240 (Note 2) 3XX austenitic stainless steel weld filler metals SFA-5.4: E308, E308L, E309, E309L, E316, E316L (Note 4) SFA-5.9: ER308, ER308L, ER309, ER309L, ER316, ER316L (Note 4) SFA-5.22: E308, E308L, #309, E309L, ER316, ER316 (Note 4) 4XX martensitic stainless steel weld filler metals SFA-5.4: E410NiMo SFA-5.9: ER410NiMo Nickel-base alloy weld filler metals SFA-5.11: ENiCrFe-7 SFA-5.14: ERNiCrFe-7, ERNiCrFe-7A, EQNiCrFe-7, EQNiCrFe-7A Notes:

1. 0.03 percent maximum carbon if unstabilized Type 3XX base metals are welded or exposed to temperature range of 800 degrees F to 1500 degrees F subsequent to final solution anneal.

2. 0.04 percent maximum carbon for FXM-19 and Type 2XX weld filler metals.

3. SB-637 UNS N07718 solution treatment temperature of range before precipitation hardening treatment restricted to 1800 degrees F to 1850 degrees F. In addition, nuts for the CNV main flange closure bolting are case-hardened to reduce galling.

4. 0.03 percent maximum carbon for unstabilized AISI Type 3XX weld filler metals; ferrite number in the range of 5FN to 20FN, except 5FN to 16FN for Type 316 and Type 316L.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 8.1.1 Pool Water Chemistry The reactor pool chemistry is maintained consistent with the spent fuel pool chemistry requirements in the Electric Power Research Institute (EPRI) Primary Water Chemistry Guidelines (Reference 9.14). The reactor pool water chemistry parameters are listed in Table 8-2. No target limit and monitoring frequency are set for lithium, hydrogen peroxide, magnesium, calcium, or aluminum. Lithium is not expected to be added in the pool and hydrogen peroxide addition is only required as needed. The control limit and monitoring requirement for calcium, magnesium, and aluminum are specified by the fuel vendor. Control limits are not included for pH and conductivity. The monitoring frequency is the only limit specified. The control limit is established for boron, rather than pH and conductivity, because the pool pH is primarily determined by boron concentration. The remainder of the NuScale reactor pool water chemistry parameters are listed in Table 8-2 with their target limits. Table 8-2 Target Limits for Reactor Pool Water Chemistry Parameter Units Value Boric acid (as boron) ppm 2000 Chloride ppm 0.15 Fluoride ppm 0.15 Sulfate ppm 0.15 Silica ppm 1.5 Aluminum ppb 80 Magnesium ppb 40 Calcium ppb 40 Turbidity NTU1 3.0 Gamma isotopic activity mCi/gram 0.001 Tritium mCi/gram Trend Note 1: NTU refers to a measure of the amount of suspended solids in a liquid. 8.1.2 Reactor Coolant System Coolant Chemistry The plant limits follow the EPRI Primary Water Chemistry Guidelines. Limits for chemical species are provided in Table 8-3. These reactor coolant chemistry specifications conform to the recommendations of RG 1.44 (Reference 9.17). The RCS water chemistry is controlled to minimize corrosion of RCS surfaces and to minimize corrosion product transport during normal operation. The CVCS provides the means for adding chemicals through charging flow and for removing chemicals through dilution or purification. For reactivity control, boric acid is added as a soluble neutron poison. The concentration of boric acid is varied throughout reactor operation as needed for reactivity control. © Copyright 2022 by NuScale Power, LLC 46

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 To maintain the alkalinity of the coolant, lithium hydroxide enriched with the lithium-7 isotope is added to the coolant. Slight alkalinity is maintained in accordance with the recommendations of the EPRI Primary Water Chemistry Guidelines. This chemical is chosen for its compatibility with boric acid, stainless steel, zirconium alloys, and nickel-base alloys. Lithium hydroxide is added to the coolant through the charging flow of the CVCS. It is removed from the coolant by the purification systems of the CVCS or reduced in concentration by dilution. The coolant pH is determined based on the recommendations in the EPRI PWR Primary Water Chemistry Guidelines and fuel vendor limits. The coolant maintains its reducing environment by adding dissolved hydrogen to the coolant. Hydrogen is used because of its compatibility with the aqueous environment and its ability to suppress radiolytic oxygen generation during normal operation. Dissolved hydrogen is added to the coolant by direct injection of high-pressure gaseous hydrogen into the CVCS charging flow. During startup, oxygen is removed by a combination of mechanical degasification by the CES and by chemical degasification using hydrazine. Hydrazine is an effective oxygen scavenger at low temperatures and is added to the coolant by the charging flow of the CVCS. Table 8-3 Reactor Coolant System Coolant Chemistry Parameter) Normal Operating Range RG 1.44 Limit Chloride (ppm) 0.15 0.15 Fluoride (ppm) 0.15 0.15 Dissolved oxygen (ppm), operating 0.10 0.10 Sulfate (ppm) 0.05 N/A Hydrogen (cc/kg), operating 15 - 50 N/A Boron (ppm) 0 - 2,000 N/A 8.2 Aging Degradation Assessment This section assesses the following aging degradations for the CNV pressure boundary materials: fatigue boric acid corrosion primary water stress corrosion-cracking (PWSCC) stress corrosion-cracking (SCC) of austenitic stainless steels SCC of pressure-retaining bolting materials irradiation embrittlement of lower CNV SCC of F6NM martensitic stainless steel © Copyright 2022 by NuScale Power, LLC 47

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 8.2.1 Fatigue Pressure-retaining components of the CNV were analyzed for fatigue in accordance with applicable subsections of ASME Code Section III. The CNV components that are part of the RCPB are described in Section 8.1. For the CNV nozzles and their safe ends that are in contact with RCS coolant during normal plant operation, the fatigue analysis considers the environmental effects in accordance with RG 1.207 (Reference 9.18) and NUREG/CR-6909 (Reference 9.19). Therefore, cracking of CNV pressure-retaining components due to fatigue loading is unlikely during the design lifetime. 8.2.2 Boric Acid Corrosion The pressure-retaining components for the CNV do not use low-alloy steel cladded or non-cladded with austenitic stainless steel. Pressure-retaining materials are either stainless steels or nickel-base alloys. Because CNV surfaces in contact with borated pool water or reactor coolant are corrosion-resistant stainless steels or nickel-base alloys, boric acid corrosion is not an applicable aging degradation mechanism for the CNV. 8.2.3 Primary Water Stress Corrosion-Cracking Alloy 600 and its weld Alloy 82/182 are susceptible to PWSCC when exposed to high-purity deaerated, hydrogenated primary water at elevated temperatures. The CNV components that are part of the RCPB are described in Section 8.1. The nickel-base alloy in contact with primary water is limited to Alloy 52/152 for the dissimilar metal welds between F304 safe-ends and the F6NM CNV top head. Extensive laboratory testing and PWR operating experience have confirmed Alloy 690/52/152 are highly resistant to PWSCC (Reference 9.7). The NuScale primary chemistry follows EPRI Primary Water Chemistry Guidelines that also minimize PWSCC. Therefore, PWSCC of Alloy 52/152 welds is unlikely. 8.2.4 Stress Corrosion-Cracking of Austenitic Stainless Steels The austenitic stainless steels for the CNV pressure-retaining components other than pressure-retaining bolting are the following: SA-965 Grade FXM-19 (UNS S20910) for lower CNV Type 2XX weld filler metals for welding SA-965 Grade FXM-19 SA-182 Grade F304 safe-ends Type 3XX weld filler metals for welding SA-182 Grade F304 safe-ends Type 3XX weld filler metals are used for welds between Type 304 safe-ends and Type 304 piping. The circumferential welds between SA-965 Grade FXM-19 components in the lower CNV are joined with Type 2XX weld filler metals. © Copyright 2022 by NuScale Power, LLC 48

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 The carbon content of SA-182 Grade F304 safe-ends and Type 3XX weld filler metals is limited to 0.03 percent maximum. The carbon content of SA-965 Grade FXM-19 and Type 2XX weld filler metals is limited to 0.04 percent maximum. The limit on carbon content minimizes intergranular precipitation Cr-carbides due to exposure to elevated temperatures during welding and post-weld heat treatment (PWHT). If water quenching is not used following final solution anneal of SA-182 Grade F304, non-sensitization is verified by the American Society for Testing and Materials (ASTM) A262 Practice A or E. The pressure-retaining austenitic stainless steels or welds are in contact with reactor pool water or RCS coolant during plant operation or NPM movement. Section 8.1 describes that the reactor pool chemistry is maintained consistent with the spent fuel pool chemistry requirements in EPRI Primary Water Chemistry Guidelines. Type 304 austenitic stainless steel has been used for spent fuel pool liners and spent fuel pool racks with excellent operating experience. Occasionally, transgranular SCC has been observed in Type 304 piping when exposed to borated water near ambient temperatures in PWRs. This observation has been attributed to sensitization, elevated chloride concentration, and high residual stresses from welding (Reference 9.8). However, transgranular SCC is unlikely to occur in CNV SA-182 Grade F304 safe-ends based on the following considerations: Sensitization is prevented by limiting carbon content to 0.03 percent maximum. Nonsensitization is verified by ASTM A262 Practice A or E in accordance with RG 1.44 as described above. The pool chemistry is maintained consistent with the spent fuel pool chemistry requirements in EPRI Primary Water Chemistry Guidelines. Deleterious species in the water are monitored to remain below acceptable limits. Halogens such as chloride and fluoride are kept to 0.15 ppm maximum. The SA-965 Grade FXM-19 (UNS S20910) is used for the lower CNV (Figure 8-1). This material is commonly referred to as XM-19 or Nitronic 50. Because of its higher yield strength and better SCC resistance than Type 304, XM-19 has been used extensively in boiling water reactor internals. According to Reference 9.9 there has been no failure or cracking of XM-19 after more than 25 years of service. 8.2.5 Stress Corrosion-Cracking of Pressure-Retaining Bolting Materials The pressure-retaining bolting materials in the CNV are: SB-637 UNS N07718 (also known as Alloy 718) SA-564 Grade 630 (also known as Type 17-4PH), H1100 SA-193 Grade B8 Class 1 and SA-194 Grade 8 (also known as Type 304) Alloy 718 is used for studs and nuts for the CNV main closure flange between the upper CNV and lower CNV, and for various closure covers in the upper CNV shell. Type 17-4PH is used for studs and nuts for various closure covers in the CNV top head. © Copyright 2022 by NuScale Power, LLC 49

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 The pressure-retaining bolting materials are in contact with the pool water only during plant operation or NPM movement. As described in Section 8.1, the pool chemistry is maintained consistent with the spent fuel pool chemistry requirements in EPRI Primary Water Chemistry Guidelines. Alloy 718 is an austenitic, precipitation-hardenable alloy whose composition is adjusted to enable strengthening by heat treatment. Alloy 718 has been used inside PWRs because of its excellent SCC resistance in primary water, although intergranular stress corrosion-cracking (IGSCC) of Alloy 718 has been reported (Reference 9.10). The Alloy 718 bolting for the CNV is submerged in the pool water during plant operation. Intergranular SCC of nickel-base alloys is unlikely at the pool water temperature of 100 degrees F. Alloy 718 contains at least 50 percent nickel. Alloys containing more than 30 percent nickel are extremely resistant to chloride-induced transgranular SCC (Reference 9.11). Therefore, SCC of Alloy 718 bolting in the CNV is unlikely. Type 17-4PH is a martensitic precipitation hardenable stainless steel. Type 17-4PH in the H900 condition is relatively susceptible to SCC. However, laboratory SCC testing showed Type 17-4PH in the overaged H1100 condition is much more resistant to SCC (Reference 9.12). Type 17-4PH for CNV bolting is used in the overaged H1100 condition. There have been no reports of SCC of Type 17-4PH in the H1100 condition in PWR applications. Failure of Type 17-4PH in the H1100 condition due to thermal embrittlement has been reported after exposure to temperatures above 500 degrees F (Reference 9.13). However, thermal aging embrittlement is not a concern because Type 17-4PH is used in CNV locations where normal operation temperature is below 500 degrees F. The pool chemistry is maintained consistent with the spent fuel pool chemistry requirements in EPRI Primary Water Chemistry Guidelines. Deleterious species in the water are monitored to remain below acceptable limits. Halogens such as chloride and fluoride are kept to 0.15 ppm maximum. Therefore, SCC of Type 17-4PH bolting in the CNV is unlikely. Type 304 austenitic stainless bolting materials (i.e., SA-193 Grade B8 and SA-194 Grade 8) are used in the solution-annealed condition for CNV pressure-retaining applications. If water quenching is not used following final solution anneal, non-sensitization is verified by ASTM A262 Practice A or E. Bolting materials are not subject to sensitization temperature range of 800 to 1500 degrees F after final solution anneal. Therefore, SCC of Type 304 bolting in the CNV is unlikely. 8.2.6 Irradiation Embrittlement of Lower Containment Vessel Figure 8-1 shows the pressure-retaining materials in the lower CNV. The lower CNV beltline region with a peak 57 effective full-power years (EFPY) fluence exceeding 1E+17 n/cm2, > 1 MeV is bounded by the lower head, lower core region shell, and lower transitional shell. To avoid irradiation embrittlement in the lower CNV beltline region, SA-965 Grade FXM-19 austenitic stainless steel and compatible Type 2XX © Copyright 2022 by NuScale Power, LLC 50

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 austenitic stainless steel weld filler metal are selected for the base metal and associated welds, respectively. The lower flange and lower shell are farther away from the core and their peak 57 EFPY fluence is below 1E+17 n/cm2, > 1 MeV. For the lower CNV non-beltline region, ASME BPVC Case N-774 SA-336 Grade F6NM martensitic stainless steel and compatible E/ER-410NiMo martensitic weld filler metal are selected for the lower flange and F6NM lower shell, and the weld between the two F6NM forgings, respectively. Because their fluence is below 1E+17 n/cm2, > 1 MeV, the materials for the lower CNV non-beltline region materials do not have an irradiation embrittlement concern. The 57 EFPY peak fluence in the lower CNV beltline region are the following: (1) CNV beltline base metal (SA-965 Grade FXM-19): 5.0E+18 n/cm2, E > 1 MeV; and (2) CNV beltline weld metal (Type 2XX): 2.5E+18 n/cm2, E > 1 MeV. Fluence is usually converted to average number of displacements per atom (dpa) to enable comparison of irradiation embrittlement data originated from different reactor types, fluence is usually converted to average number of displacements per atom (dpa). A typical conversion factor for light water reactors is 1dpa = 6.7E+20 n/cm2 per MRP-175 (Reference 9.20). Using this conversion factor, the dpa equivalent of the lower CNV peak fluence is the following: (1) CNV beltline base metal: 5.0E+18 n/cm2, E > 1 MeV = 0.0075 dpa; and (2) CNV beltline weld metal: 2.5E+18 n/cm2, E > 1 MeV

             = 0.0037 dpa.

Based on extensive irradiated Type 3XX fracture toughness data, MRP-175 proposed the following screening fluence for irradiation embrittlement in PWR reactor internals: (1) wrought austenitic stainless steels > 1.5 dpa; and (2) austenitic stainless steel welds or cast austenitic stainless steel > 1 dpa. However, NUREG/CR-7027 (Reference 9.15) proposed the following more conservative threshold fluence for irradiation embrittlement in Type 3XX austenitic stainless steels than the MRP-175 screening fluence: (1) wrought austenitic stainless steels = 0.5 dpa; and (2) austenitic stainless steel welds or cast austenitic stainless steel = 0.3 dpa. Although XM-19 and E209/ER209 or E240/ER240 welds contain higher manganese and nitrogen content than Type 3XX, there is no data to indicate such differences have a pronounced effect on irradiation embrittlement. Fracture toughness testing of irradiated solution annealed XM-19 was performed under an EPRI- Department of Energy program (Reference 9.16). After 0.28 dpa at 340 degrees C (644 degrees F), fracture toughness was found to be JIc = 198 kJ/m2 (KJc = 212 MPavm) when tested at 289 degrees C (552 degrees F). This XM-19 test data point is well within the scatter band of Type 3XX austenitic stainless steels reviewed by MRP-175 or NUREG/CR-7027. Therefore, XM-19 irradiation embrittlement behavior is similar to Type 3XX, at least up to 0.28 dpa. © Copyright 2022 by NuScale Power, LLC 51

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 The peak 57 EFPY fluence is 0.0075 dpa for the lower CNV beltline base metal and 0.0035 dpa for the lower CNV beltline welds. These peak fluence values are tiny fractions of either MRP-175 screening fluence or NUREG/CR-7027 threshold fluence for irradiation embrittlement. Therefore, loss of fracture toughness in the lower CNV beltline SA-965 Grade FXM-19 base metal or associated weld metal from neutron irradiation during the design lifetime is negligible. Figure 8-1 Lower Containment Vessel Pressure-Retaining Materials 8.2.7 Stress Corrosion-Cracking of F6NM Martensitic Stainless Steel Except for the beltline of the lower CNV, which is made of austenitic stainless steel, the CNV (top head, top head cover, nozzles, access ports, covers for access ports, and flanges) is made of Grade F6NM martensitic stainless steel per ASME BPVC Case N-774 or per SA-182 (Table 8-1). The F6NM base metal is tempered at 1095 +/- 25 degrees F for 8 hours in two steps of 4 hours each. After each tempering step, the base metal is cooled to below 175 degrees F. F6NM-to-F6NM welds are made with 4XX weld filler metals in Table 8-1. For dissimilar metal welds between F6NM and FXM-19 and between F6NM and Type 304, the F6NM base metal is buttered with 2XX weld filler metals or with nickel-base alloy weld filler metals in Table 8-1, respectively. The PWHT after welding of F6NM-to-F6NM or after buttering of F6NM base metal is in accordance with ASME © Copyright 2022 by NuScale Power, LLC 52

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 BPVC, Section III, NB-4622 for P-No.6 Gr.4. The PWHT temperature is 1075 +/- 25 degrees F for minimum holding time per ASME BPVC, Section III, Table NB-4622.1-1. The CRDM pressure housings are known examples of welded martensitic stainless steel pressure-retaining components in LLWRs. Type 403 martensitic stainless steel has been used in CRDM pressure housing of United States PWRs since the 1970s. In such CRDM designs, the Type 403 section was welded to austenitic stainless steel by dissimilar metal welds to form a complete pressure housing. There have been no reports of degradation of CRDM pressure housings containing Type 403, except for one leaking CRDM pressure housing at Prairie Island Unit 2 in 1998. Subsequent failure analysis (Reference 9.21) identified a pre-existing crack from fabrication and concluded the leak was due to a fabrication defect. Alloy F6NM martensitic stainless steel has been used in CRDM pressure housings of German Konvoi PWRs since 1988. The Konvoi CRDM design was later adopted by US EPR. According AREVAs responses to NRC requests for additional information (RAIs) (Reference 9.22 and Reference 9.23), the Konvoi CRDM pressure housing had been in service for 19 years (at the time of AREVA response to RAIs in 2009) without crack indications or leakages. The upper dissimilar metal weld was inspected by eddy-current examination from the inside diameter, and by ultrasonic examination and liquid penetrant examination from the outside diameter. No intergranular corrosion attack, cracking due to IGSCC, transgranular SCC, and thermal embrittlement, or leakages had been detected in any Konvoi CRDM pressure housings. The Konvoi PWRs in Germany are still operating as of October 2022, but are scheduled for permanent shutdown by the end of 2022. They have operated for about 34 years since 1988 without reports of cracking or degradation issues related to the CRDM pressure housings and the dissimilar metal welds. In the response to NRC RAIs (Reference 9.22 and Reference 9.23), AREVA stated that F6NM is not susceptible to SCC based on laboratory SCC testing at 599 degrees F. The specimens were loaded to 90 percent of room temperature yield strength. No cracking was observed in any of the 177 specimens tested under simulated PWR primary water after 4200 hours. Additional testing was performed under more aggressive water chemistry:

a. 1750 ppm H3BO3 + 10 ppm chloride/N2H4 for 880 hours

b. 1750 ppm H3BO3 + 100 ppm chloride/N2H4 for 984 hours

c. Oxygen saturated water containing 100 ppm chloride No cracking was observed under conditions (a) and (b), but cracking was observed under condition (c). However, AREVA noted condition (c) was far outside normal PWR primary water chemistry.

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 The design ensures the CNV inside surface is exposed to vacuum during operation. The top portion of the upper CNV is not submerged during plant operation. The rest of the CNV outside surface is in contact with reactor pool water, whose water chemistry is similar to PWR primary water during shutdown period (Section 8.1). Type 410 martensitic stainless steel is susceptible to SCC in LLWRs if it is tempered below 1050 degrees F, resulting in hardness in excess of Rockwell C hardness number (HRC) 33. Some examples of SCC of Type 410 are listed below. The NRC Information Notice No. 85-59 (Reference 9.24) described several incidents of cracked Type 410 valve stems and shafts due to IGSCC. In all cases, the cracking was attributed to excessive hardness levels corresponding to tempering temperature between 700 to 1050 degrees F. The NRC Information Notice No. 86-39 (Reference 9.25) described cracked Type 410 wear rings in residual heat removal pumps due to IGSCC. The cracking was attributed to excessive hardness HRC 33 to 39. The NRC Information Notice No. 88-85 (Reference 9.26) described cracked Type 410 studs due to IGSCC. The cracking was attributed to excessive hardness HRC 36. The NRC Bulletin 89-02 (Reference 9.27) described cracked Type 410 bolting for a swing check valve in the residual heat removal system. Cracking was due to SCC and was attributed to excessive hardness HRC 36. It was not reported if the SCC was IGSCC or transgranular SCC. The NRC Information Notice No. 95-26 (Reference 9.28) described cracked Type 416 lock nuts for safety-injection pumps due to IGSCC. The cracking was attributed to excessive hardness HRC 47. Type 416 is the free-machining grade of Type 410 (intentionally adding 0.15 percent minimum sulfur). Tsubota (Reference 9.29) performed laboratory SCC tests of different martensitic stainless steels including F6NM using creviced bent beam type specimens. To examine SCC sensitivity to tempering temperature, each material was tempered at several temperatures. Alloy F6NM tempered at 580 and 600 degrees C (1076 and 1112 degrees F) is consistent with the F6NM used for the CNV. The SCC testing was performed in 550-degrees F water saturated with oxygen at room temperature. After 500 hours, the creviced bent beam specimens were examined for crack depth, which was plotted as a function of hardness. Figure 8-2 shows an abrupt increase in SCC susceptibility when hardness exceeded 340 Vickers Hardness number (HV) for the materials tested. The SCC susceptibility is low if F6NM hardness level is kept below 340 HV, which is equivalent to HRC 34. The laboratory test results were also consistent with inservice SCC failures of insufficiently tempered Type 410 in LLWRs. The hardness control is confirmed to be effective by Konvoi CRDM latch housings made of F6NM. During the Konvoi CRDM housing production, the F6NM latch housing hardness was limited to 350 HV maximum (Reference 9.22 and Reference 9.23). In the case of NuScale, the maximum hardness for F6NM base © Copyright 2022 by NuScale Power, LLC 54

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 metal for the CNV is limited to Brinell hardness number 295 maximum by ASME BPVC Case N-774 and SA-182. Brinell hardness number 295 is equivalent to 310 HV or HRC 31. In addition, the minimum PWHT temperature for F6NM-to-F6NM welds and buttering of F6NM in the CNV is 1050 degrees F in accordance with ASME Section III, NB-4622. Therefore, based on the operating experience of welded CRDM pressure housings containing martensitic stainless steels and laboratory SCC test results of martensitic stainless steels in simulated PWR and BWR primary water, SCC of Grade F6NM used in the CNV is unlikely. Figure 8-2 Stress Corrosion Cracking Depth as a Function of Hardness, Martensitic Stainless Steels (Reference 9.29) © Copyright 2022 by NuScale Power, LLC 55

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 9.0 References 9.1 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III Division 1, Rules for Construction of Nuclear Facility Components, 2017 edition, New York, NY. 9.2 American Society of Mechanical Engineers, Operation and Maintenance of Nuclear Power Plants, ASME OM-2017, New York, NY. 9.3 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section XI Division 1, Rules for Inservice Inspection of Nuclear Power Plant Components, 2017 edition, New York, NY. 9.4 American National Standards Institute/American Nuclear Society, Containment System Leakage Testing Requirements, ANSI/ANS 56.8, 1994, La Grange Park, IL. 9.5 Institute of Electrical and Electronics Engineers, IEEE Standard for Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations, IEEE Standard 317-1983 (R2003), New York, NY. 9.6 Nuclear Energy Institute, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, NEI 94-01, Rev. 3-A, July 2012. 9.7 Electric Power Research Institute, Materials Reliability Program: Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111), EPRI #1009801, 2004, Palo Alto, CA. 9.8 NRC Information Notice 2011-04: Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized Water Reactors. 9.9 Electric Power Research Institute, EPRI Materials Management Matrix Project: Economic Simplified Boiling Water Reactor Degradation Matrix, Rev. 0, EPRI

                  #1016332, 2008, Palo Alto, CA.

9.10 McIlree, A. R., Degradation of High Strength Austenitic Alloys X-750, 718 and A286 in Nuclear Power Systems, 1st International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, NACE, 1984. 9.11 U.S. Nuclear Regulatory Commission, A Review of Stress Corrosion Cracking of High-Level Nuclear Waste Container Materials, CNWRA 92-021, August 1992. 9.12 Rowland, M. C. and W. R. Smith, Sr., Precipitation-Hardening Stainless Steels in Water-Cooled Reactors, Nuclear Engineering, (January 1962), pp. 14-22. © Copyright 2022 by NuScale Power, LLC 56

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 9.13 Xu, H. and S. Fyfitch, Aging Embrittlement Modeling of Type 17-4 PH at LWR Temperature, Proceedings of the 10th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, August 3-9, 2001. 9.14 Electric Power Research Institute, Pressurized Water Reactor Primary Water Chemistry Guidelines, Technical Report 3002000505, Rev. 7, April 2014, Palo Alto, CA. 9.15 U.S. Nuclear Regulatory Commission, Degradation of LWR Core Internal Materials due to Neutron Irradiation, NUREG/CR-7027, December 2010. 9.16 Teysseyre, S., et al., Irradiation Assisted Stress Corrosion Cracking Susceptibility of Alloy X-750 and XM-19 Exposed to BWR Environments, Presentation at International Light Water Reactor Materials Reliability Conference and Exhibition, August 1-4, 2016, Chicago, Illinois. 9.17 U.S. Nuclear Regulatory Commission, Control of the Processing and Use of Stainless Steel, Regulatory Guide 1.44, Rev. 1, March 2011. 9.18 U.S. Nuclear Regulatory Commission, Guidelines for Evaluating Fatigue Analyses Incorporating the Life Reduction of Metal Components Due to the Effects of the Light-Water Reactor Environment for New Reactors, Regulatory Guide 1.207, March 2007. 9.19 U.S. Nuclear Regulatory Commission, Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials, NUREG/CR-6909, Rev. 1, Draft Report for Comment, March 2014. 9.20 Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175). EPRI, Palo Alto, CA: 2005. 1012081. 9.21 WCAP-15054, Metallurgical Investigation and Root Cause Assessment of Part Length CRDM Housing Motor Tube Cracking at Prairie Island Nuclear Generating Plant Unit 2 9.22 Response to US EPR Design Certification Application RAI No. 199, Supplement 1, NRC Accession Number ML091560436. 9.23 Response to US EPR Design Certification Application RAI No. 199, Supplement 2, NRC Accession Number ML101310011. 9.24 NRC Information Notice No. 85-59: Valve Stem Corrosion Failures. 9.25 NRC Information Notice No. 86-39: Failures of RHR Pump Motors and Pump Internals. © Copyright 2022 by NuScale Power, LLC 57

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 9.26 NRC Information Notice No. 88-85: Broken Retaining Block Studs on Anchor Darling Check Valves. 9.27 NRC Bulletin 89-02: Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Darling Model S350w Swing Check Valves or Valves of Similar Design. 9.28 NRC Information Notice No. 95-26: Defect In Safety-Related Pump Parts Due To Inadequate Heat Treatment. 9.29 M. Tsubota, et al., Effect of Tempering on SCC Susceptibility of Martensitic Stainless Steels in High Temperature Water, Proceedings of the 4th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, NACE, 1989, pp. 9-66 through 9-75. © Copyright 2022 by NuScale Power, LLC 58

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Appendix A Containment Isolation Summary Figures The following figures are provided to show the isolation valves and closed loop piping systems that form part of the containment pressure boundary for the GDC 55, GDC 56 and GDC 57 piping penetrations of the CNV. Collectively, these figures identify the fluid service penetrations of the NuScale containment. Table A-1 Simplified Figures Illustrating the Containment Pressure Boundary for the General Design Criteria 55, General Design Criteria 56 and General Design Criteria 57 Piping Systems of the Containment Vessel. A-1 CNTS A-2 DHRS A-3 SGS © Copyright 2022 by NuScale Power, LLC A-1

Table B-1 Type B Containment Penetrations

     © Copyright 2022 by NuScale Power, LLC Location     Quantity of    Type B Nominal Size Pen.    [Type] Component                                  (Azimuth,    EPA Sheath    Leakage   Notes (Opening)

Elevation) Modules Test? CNV8 [EPA] Instrumentation and controls NPS 3 348°, 702 3 blanks 4 Yes (I&C) Division 1 CNV9 [EPA] I&C Division 2 NPS 3 12°, 702 4 Yes 3 blanks CNV15 [EPA] PZR Heater Power 1 NPS 12 235°, 702 9 Yes 3 blanks CNV16 [EPA] PZR Heater Power 2 NPS 12 125°, 702 9 Yes 3 blanks CNV17 [ISA] ICI Channel A 8 inch CNV Head N/A Yes 3 ICI points, 1 blank CNV18 [ISA] ICI Channel C 8 inch CNV Head N/A Yes 3 ICI points, 1 blank CNV19 [ISA] ICI Channel B 8 inch CNV Head N/A Yes 3 ICI points, 1 blank CNV20 [ISA] ICI Channel D 8 inch CNV Head N/A Yes 3 ICI points, 1 blank CNV24 [Port] CNV Manway Access 1 NPS 18 315°, 719 N/A Yes CNV25 [Port] CRDM Access 67 inch CNV Head N/A Yes CNV26 [Port] CNV Manway Access 2 38 inch 315°, 719 N/A Yes CNV27 [Port] SG Access 1 38 inch 45°, 576 N/A Yes CNV28 [Port] SG Access 2 38 inch 315°, 576 N/A Yes CNV29 [Port] SG Access 3 38 inch 225°, 576 N/A Yes CNV30 [Port] SG Access 4 38 inch 135°, 576 N/A Yes CNV31 [Port] PZR Heater Access 1 44 inch 90°, 596 N/A Yes CNV32 [Port] PZR Heater Access 2 44 inch 270°, 596 N/A Yes CNV33 353°, 585 Notes 3 and 4 NuScale Containment Leakage Integrity Assurance [Valve] RVV Trip/Reset A(3) NPS 3(4) N/A Yes CNV34 [Valve] RVV Trip/Reset B(3) NPS 3(4) 7°, 585 N/A Yes Notes 3 and 4 CNV35 [Valve] RRV Trip/Reset A(3) NPS 3(4) 353°, 408 N/A Yes Notes 3 and 4 CNV36 [Valve] RRV Trip/Reset B(3) NPS 3(4) 7°, 408 N/A Yes Notes 3 and 4 CNV37 [EPA] CRDM Power 1 NPS 18 135°, 742 9 Yes 1 blank CNV38 [EPA] RPI Group 1 NPS 10 105°, 742 4 Yes No blanks CNV39 [EPA] RPI Group 2 NPS 10 270°, 742 4 Yes No blanks CNV40 [EPA] I&C Separation Group A NPS 8 348°, 727 4 Yes 1 blank CNV41 [EPA] I&C Separation Group B NPS 8 12°, 727 4 Yes No blanks TR-123952-NP CNV42 [EPA] I&C Separation Group C NPS 8 73°, 742 4 Yes No blanks CNV43 [EPA] I&C Separation Group D NPS 8 287°, 742 4 Yes 1 blank CNV44 [EPA] CRDM Power 2 NPS 18 225°, 742 9 Yes 1 blank B-2 N/A [N/A] CNV Closure Flange 170 inch 328.3 N/A Yes Revision 0

Table B-1 Type B Containment Penetrations (Continued)

     © Copyright 2022 by NuScale Power, LLC Location        Quantity of       Type B Nominal Size Pen.          [Type] Component                                    (Azimuth,       EPA Sheath       Leakage     Notes (Opening)

Elevation) Modules Test? CNV5 [CITF] RCCW-HV-0905(5) NPS 2 CNV Head N/A Yes CITF cover only CNV6 [CITF] CVC-HV-0906 (5) NPS 2 CNV Head N/A Yes CITF cover only CNV7 [CITF] CVC-HV-0907 (5) NPS 2 CNV Head N/A Yes CITF cover only CNV10 [CITF] CE-HV-0910(5) NPS 4 CNV Head N/A Yes CITF cover only CNV11 [CITF] CFD-HV-0911(5) NPS 2 CNV Head N/A Yes CITF cover only CNV12 [CITF] RCCW-HV-0912(5) NPS 2 CNV Head N/A Yes CITF cover only CNV13 [CITF] CVC-HV-0913(5) NPS 2 CNV Head N/A Yes CITF cover only CNV14 [CITF] CVC-HV-0914(5) NPS 2 CNV Head N/A Yes CITF cover only Notes:

1. Penetration ID number CNV21 is not used.

2. Elevations are measured from global zero (bottom of the CNV support skirt) measured to the top of the respective safe-end or nozzle cover.

3. RVV and RRV trip/reset valves are part of the reactor coolant pressure boundary.

4. CNV33-36 are 3-inch penetrations for ECCS trip and reset valves. Each penetration has two bolted connections (trip and reset valve) that each require a Type B test at the body-to-bonnet joint.

5. Each CITF is common to both CIVs in a given penetration.

TR-123952-NP NuScale Containment Leakage Integrity Assurance B-3 Revision 0

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Appendix C List of Type C Containment Penetrations Table C-1 Penetration CNV1 Penetration CNV1 Category Parameter Valve FW-HOV-0137 Valve SG-RV-0102 Feedwater Isolation Valve SGS thermal relief valve Name (FWIV) #1 Location CNV head FW line inside CNV Valve Type Ball valve Spring-actuated relief valve Hydraulic to open, stored Spring close, line pressure to Operator Type energy device to close open Overview Fluid Water Water Nominal Size (Opening) NPS 4 (inlet), NPS 5 (outlet) NPS 3/4 x 1 GDC 57 Connected to GDC 57 closed CIV Configuration loop No, TS leak test per IST No, TS leak test per IST Type C Leakage Test? program program Normal Open Closed Shutdown Closed Closed Positions Safety Function Closed Open Failure Closed N/A Primary Automatic Self-actuating Secondary Remote manual N/A Actuation Hydraulic actuator with failsafe Spring Power Source stored energy device Design Pressure 2200 psia 2200 psia Design Temperature 650 degrees F 650 degrees F Design Seismic Category I I Design Code Valve Section III, NC Section III, NC Table C-2 Penetration CNV2 Penetration CNV2 Category Parameter Valve FW-HOV-0237 Valve SG-RV-0202 Feedwater Isolation Valve SGS thermal relief valve Name (FWIV) #2 Location CNV head FW line inside CNV Valve Type Ball Valve Spring-actuated relief valve Hydraulic to open, stored Spring close, line pressure to Operator Type energy device to close open Overview Fluid Water Water Nominal Size (Opening) NPS 4 (inlet), NPS 5 (outlet) NPS 3/4 x 1 GDC 57 Connected to GDC 57 closed CIV Configuration loop No, TS leak test per IST No, TS leak test per IST Type C Leakage Test? program program © Copyright 2022 by NuScale Power, LLC C-1

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-2 Penetration CNV2 (Continued) Penetration CNV2 Category Parameter Valve FW-HOV-0237 Valve SG-RV-0202 Normal Open Closed Shutdown Closed Closed Positions Safety Function Closed Open Failure Closed N/A Primary Automatic Self-actuating Secondary Remote manual N/A Actuation Hydraulic actuator with failsafe Spring Power Source stored energy device Design Pressure 2200 psia 2200 psia Design Temperature 650 degrees F 650 degrees F Design Seismic Category I I Design Code Valve Section III, NC Section III, NC © Copyright 2022 by NuScale Power, LLC C-2

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-3 Penetration CNV3 Penetration CNV3 Category Parameter Valve MS-HOV-101 Valve MS-HOV-103 Main Steam Isolation Valve Main Steam Isolation Bypass Name (MSIV) #1 Valve (MSIBV) #1 Location CNV head CNV head, parallel to MSIV Valve Type Ball valve Ball valve Hydraulic to open, stored Hydraulic to open, stored Operator Type Overview energy device to close energy device to close Fluid Steam Steam Nominal Size (Opening) NPS 12 NPS 2 CIV Configuration GDC 57 GDC 57 No, TS leak test per IST No, TS leak test per IST Type C Leakage Test? program program Normal Open Open Shutdown Closed Closed Positions Safety Function Closed Closed Failure Closed Closed Primary Automatic Automatic Secondary Remote manual Remote manual Actuation Hydraulic actuator with failsafe Hydraulic actuator with failsafe Power Source stored energy device stored energy device Design Pressure 2200 psia 2200 psia Design Temperature 650 degrees F 650 degrees F Design Seismic Category I I Design Code Valve Section III, NC Section III, NC Notes: (1) DHRS penetration CNV45 is integral to CNV3. © Copyright 2022 by NuScale Power, LLC C-3

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-4 Penetration CNV4 Penetration CNV4 Category Parameter Valve MS-HOV-0201 Valve MS-HOV-0203 Main Steam Isolation Valve Main Steam Isolation Bypass Name (MSIV) #2 Valve (MSIBV) #2 Location CNV head CNV head, parallel to MSIV Valve Type Ball valve Ball valve Hydraulic to open, stored Hydraulic to open, stored Operator Type Overview energy device to close energy device to close Fluid Steam Steam Nominal Size (Opening) NPS 12 NPS 2 CIV Configuration GDC 57 GDC 57 No, TS leak test per IST No, TS leak test per IST Type C Leakage Test? program program Normal Open Open Shutdown Closed Closed Positions Safety Function Closed Closed Failure Closed Closed Primary Automatic Automatic Secondary Remote manual Remote manual Actuation Hydraulic actuator with failsafe Hydraulic actuator with failsafe Power Source stored energy device stored energy device Design Pressure 2200 psia 2200 psia Design Temperature 650 degrees F 650 degrees F Design Seismic Category I I Design Code Valve Section III, NC Section III, NC Notes: (1)DHRS penetration CNV46 is integral to CNV4. © Copyright 2022 by NuScale Power, LLC C-4

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-5 Penetration CNV5 Penetration CNV5 Category Parameter Valve RCCW-HOV-0190 Valve RCCW-HOV-0191 RCCWS Return Containment RCCWS Return Containment Name Isolation Valve, Inboard Isolation Valve, Outboard Location CNV head, CNV side CNV head, piping side Valve Type Ball valve Ball valve Hydraulic to open, stored Hydraulic to open, stored Overview Operator Type energy device to close energy device to close Fluid Water Water Nominal Size (Opening) NPS 2 NPS 2 CIV Configuration GDC 56 GDC 56 Type C Leakage Test? Yes Yes Normal Open Open Shutdown Closed Closed Positions Safety Function Closed Closed Failure Closed Closed Primary Automatic Automatic Secondary Remote manual Remote manual Actuation Hydraulic actuator with failsafe Hydraulic actuator with failsafe Power Source stored energy device stored energy device Design Pressure 2850 psia(1) 2850 psia(1) Design Temperature 650 degrees F(1) 650 degrees F(1) Design Seismic Category I I Design Code Valve Section III, NB(1) Section III, NB(1) Notes: (1) Valve provides a containment boundary and is classified as ASME BPVC Class 2 with a minimum design pressure and temperature requirement equivalent to the CNTS. However, primary system CIVs are designed to ASME BPVC Section III NB with a design pressure and temperature requirement equivalent to the CVCS injection piping (2850 psia, 650 degrees F). © Copyright 2022 by NuScale Power, LLC C-5

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-6 Penetration CNV6 Penetration CNV6 Category Parameter Valve CVC-HOV-0331 Valve CVC-HOV-0330 CVC Injection Containment CVC Injection Containment Name Isolation Valve, Inboard Isolation Valve, Outboard Location CNV head, CNV side CNV head, piping side Valve Type Ball valve Ball valve Hydraulic to open, stored Hydraulic to open, stored Overview Operator Type energy device to close energy device to close Fluid Water Water Nominal Size (Opening) NPS 2 NPS 2 CIV Configuration GDC 55 GDC 55 Type C Leakage Test? Yes Yes Normal Open Open Shutdown Closed Closed Positions Safety Function Closed Closed Failure Closed Closed Primary Automatic Automatic Secondary Remote manual Remote manual Actuation Hydraulic actuator with failsafe Hydraulic actuator with failsafe Power Source stored energy device stored energy device Design Pressure 2850 psia 2850 psia Design Temperature 650 degrees F 650 degrees F Design Seismic Category I I Design Code Valve Section III, NB Section III, NB © Copyright 2022 by NuScale Power, LLC C-6

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-7 Penetration CNV7 Penetration CNV7 Category Parameter Valve CVC-HOV-0325 Valve CVC-HOV-0324 PZR Spray Containment PZR Spray Containment Name Isolation Valve, Inboard Isolation Valve, Outboard Location CNV head, CNV side CNV head, piping side Valve Type Ball Valve Ball Valve Hydraulic to open, stored Hydraulic to open, stored Overview Operator Type energy device to close energy device to close Fluid Water Water Nominal Size (Opening) NPS 2 NPS 2 CIV Configuration GDC 55 GDC 55 Type C Leakage Test? Yes Yes Normal Open Open Shutdown Closed Closed Positions Safety Function Closed Closed Failure Closed Closed Primary Automatic Automatic Secondary Remote manual Remote manual Actuation Hydraulic actuator with failsafe Hydraulic actuator with failsafe Power Source stored energy device stored energy device Design Pressure 2850 psia 2850 psia Design Temperature 650 degrees F 650 degrees F Design Seismic Category I I Design Code Valve Section III, NB Section III, NB © Copyright 2022 by NuScale Power, LLC C-7

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-8 Penetration CNV10 Penetration CNV10 Category Parameter Valve CE-HOV-0001 Valve CE-HOV-0002 Containment Evacuation Containment Evacuation Name Containment Isolation Valve, Containment Isolation Valve, Inboard Outboard Location CNV head, CNV side CNV head, piping side Valve Type Ball Valve Ball Valve Overview Hydraulic to open, stored Hydraulic to open, stored Operator Type energy device to close energy device to close Fluid Steam and air Steam and air Nominal Size (Opening) NPS 4 NPS 4 CIV Configuration GDC 56 GDC Type C Leakage Test? Yes Yes Normal Open Open Shutdown Closed Closed Positions Safety Function Closed Closed Failure Closed Closed Primary Automatic Automatic Secondary Remote manual Remote manual Actuation Hydraulic actuator with failsafe Hydraulic actuator with failsafe Power Source stored energy device stored energy device Design Pressure 1200 psia 1200 psia Design Temperature 600 degrees F 600 degrees F Design Seismic Category I I Design Code Valve Section III, NC Section III, NC © Copyright 2022 by NuScale Power, LLC C-8

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-9 Penetration CNV11 Penetration CNV11 Category Parameter Valve CFD-HOV-0022 Valve CFD-HOV-0021 Containment Flooding Containment Flooding Name Containment Isolation Valve, Containment Isolation Valve, Inboard Outboard Location CNV head, CNV side CNV head, piping side Valve Type Ball valve Ball valve Overview Hydraulic to open, stored Hydraulic to open, stored Operator Type energy device to close energy device to close Fluid Water Water Nominal Size (Opening) NPS 2 NPS 2 CIV Configuration GDC 56 GDC 56 Type C Leakage Test? Yes Yes Normal Closed Closed Shutdown Closed Closed Positions Safety Function Closed Closed Failure Closed Closed Primary Automatic Automatic Secondary Remote manual Remote manual Actuation Hydraulic actuator with failsafe Hydraulic actuator with failsafe Power Source stored energy device stored energy device Design Pressure 2850 psia(1) 2850 psia(1) Design Temperature 650 degrees F(1) 650 degrees F(1) Design Seismic Category I I Design Code Valve Section III, NB(1) Section III, NB(1) Notes: (1) Valve provides a containment boundary and is classified as ASME BPVC Class 2 with a minimum design pressure and temperature requirement equivalent to the CNTS. However, primary system CIVs are designed to ASME BPVC Section III NB with a design pressure and temperature requirement equivalent to the CVCS injection piping (2850 psia, 650 degrees F). © Copyright 2022 by NuScale Power, LLC C-9

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-10 Penetration CNV12 Penetration CNV12 Category Parameter Valve RCCW-HOV-0185 Valve RCCW-HOV-0184 RCCWS Supply Containment RCCWS Supply Containment Name Isolation Valve, Inboard Isolation Valve, Outboard Location CNV head, CNV side CNV head, piping side Valve Type Ball valve Ball valve Hydraulic to open, stored Hydraulic to open, stored Overview Operator Type energy device to close energy device to close Fluid Water Water Nominal Size (Opening) NPS 2 NPS 2 CIV Configuration GDC 56 GDC 56 Type C Leakage Test? Yes Yes Normal Open Open Shutdown Closed Closed Positions Safety Function Closed Closed Failure Closed Closed Primary Automatic Automatic Secondary Remote manual Remote manual Actuation Hydraulic actuator with failsafe Hydraulic actuator with failsafe Power Source stored energy device stored energy device Design Pressure 2850 psia(1) 2850 psia(1) Design Temperature 650 degrees F(1) 650 degrees F(1) Design Seismic Category I I Design Code Valve Section III, NB(1) Section III, NB(1) Notes: (1) Valve provides a containment boundary and is classified as ASME BPVC Class 2 with a minimum design pressure and temperature requirement equivalent to the CNTS. However, primary system CIVs are designed to ASME BPVC Section III NB with a design pressure and temperature requirement equivalent to the CVCS injection piping (2850 psia, 650 degrees F). © Copyright 2022 by NuScale Power, LLC C-10

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-11 Penetration CNV13 Penetration CNV13 Category Parameter Valve CVC-HOV-0334 Valve CVC-HOV-0335 CVC Discharge Containment CVC Discharge Containment Name Isolation Valve, Inboard Isolation Valve, Outboard Location CNV head, CNV side CNV head, piping side Valve Type Ball valve Ball valve Hydraulic to open, stored Hydraulic to open, stored Overview Operator Type energy device to close energy device to close Fluid Water Water Nominal Size (Opening) NPS 2 NPS 2 CIV Configuration GDC 55 GDC 55 Type C Leakage Test? Yes Yes Normal Open Open Shutdown Closed Closed Positions Safety Function Closed Closed Failure Closed Closed Primary Automatic Automatic Secondary Remote manual Remote manual Actuation Hydraulic actuator with failsafe Hydraulic actuator with failsafe Power Source stored energy device stored energy device Design Pressure 2850 psia 2850 psia Design Temperature 650 degrees F 650 degrees F Design Seismic Category I I Design Code Valve Section III, NB Section III, NB © Copyright 2022 by NuScale Power, LLC C-11

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-12 Penetration CNV14 Penetration CNV14 Category Parameter Valve CVC-HOV-0401 Valve CVC-HOV-0402 RPV High Point Degasification RPV High Point Degasification Name Containment Isolation Valve, Containment Isolation Valve, Inboard Outboard Location CNV head, CNV side CNV head, piping side Valve Type Ball valve Ball valve Overview Hydraulic to open, stored Hydraulic to open, stored Operator Type energy device to close energy device to close Fluid Steam and air Steam and air Nominal Size (Opening) NPS 2 NPS 2 CIV Configuration GDC 55 GDC 55 Type C Leakage Test? Yes Yes Normal Closed Closed Shutdown Closed Closed Positions Safety Function Closed Closed Failure Closed Closed Primary Automatic Automatic Secondary Remote manual Remote manual Actuation Hydraulic actuator with failsafe Hydraulic actuator with failsafe Power Source stored energy device stored energy device Design Pressure 2850 psia 2850 psia Design Temperature 650 degrees F 650 degrees F Design Seismic Category I I Design Code Valve Section III, NB Section III, NB © Copyright 2022 by NuScale Power, LLC C-12

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Table C-13 Penetrations CNV22 and CNV23 Penetration CNV22 Penetration CNV23 Category Parameter (no valves) (no valves) Decay Heat Removal System Decay Heat Removal System Name Train 1 Condensate Penetration Train 2 Condensate Penetration Location N/A N/A Valve Type N/A N/A Overview Operator Type N/A N/A Fluid Water Water Nominal Size (Opening) NPS 2 NPS 2 CIV Configuration GDC 57(1) GDC 57(1) Type C Leakage Test? N/A N/A Normal N/A N/A Shutdown N/A N/A Positions Safety Function N/A N/A Failure N/A N/A Primary N/A N/A Actuation Secondary N/A N/A Power Source N/A N/A Design Pressure 2200 psia 2200 psia Design Temperature 650 degrees F 650 degrees F Design Seismic Category I I Design Code Valve Section III, NC Section III, NC Notes: (1) The DHRS lines have the attributes of both a closed loop inside and outside of containment. An exemption is provided to clarify the system design within GDC 57 criteria. CNV45 and CNV46 (which are integral to CNV3 and CNV4, respectively) are the DHRS lines that penetrate containment. The DHRS becomes a closed system outside containment when the FWIVs and MSIVs shut, creating the DHRS boundary. The test for this system is the leakage test of the FWIVs, MSIVs, and main steam bypass valves in accordance with the IST Program. © Copyright 2022 by NuScale Power, LLC C-13

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Appendix D Type A Testing Challenges D.1 Overview The design supports an exemption from the requirements of GDC 52 and 10 CFR 50, Appendix J, Type A tests, which specify the design for and performance of preoperational and periodic ILRT at containment design pressure. However, NuScale reviewed the requirements of GDC 52 and 10 CFR 50, Appendix J, Type A testing to assess the potential of performing ILRT with the NuScale design. The inherent safety feature of a small metal containment in direct contact with the UHS presents unique challenges to performing ILRT with the NuScale design. At the conclusion of a normal shutdown for refueling, the CNV is filled with water to provide heat transfer during reactor refueling by filling the containment with water up to a level near the reactor PZR baffle plate. The heat transfer across the reactor vessel wall into the containment filled with water and through the containment wall into the UHS water provides cooling for the fuel in the RPV (Figure D-1). The high heat transfer ability of the system coupled with the changing decay heat from the core, as well as the UHS heat transfer coupled to the rest of the NPMs in the UHS pool, creates a highly variable temperature system. D.2 Temperature To ensure temperature variations are detected and offset, high-precision sensors both in the top of the PZR and in the containment gas space are provided (Figure D-1). If the RPV water level were lower than the baffle plate, then the additional area under the baffle would need to be individually instrumented. Sensors needed to monitor temperature changes of the coolant in the RPV and the CNV would need to be more accurate and placed in different locations than the normal plant temperature sensors inside the RPV and the CNV. While the exact number of additional sensors required for ILRT is unknown, including these sensors permanently in the design would complicate CNV instrumentation and add more signal leads to those already required. Additionally, permanently installed sensors may not be in the optimal locations for a given test. Differing conditions (e.g., water level, air and water temperature) than those for the "design" test requires review and possible reconfiguration of instrument quantities and locations to provide meaningful ILRT test results. © Copyright 2022 by NuScale Power, LLC D-1

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Figure D-1 Reactor Pressure Vessel, Containment Vessel, and Ultimate Heat Sink Temperature Gradients D.3 Temperature Changes During ILRT, a coincident undetected temperature change of the gas volume would result in an uncompensated change in pressure of magnitude similar to the allowable pressure change associated with the test leak rate limit. From the Combined Gas Law, a 0.1 degree F temperature rise increases CNV pressure proportionately: P1 T

             ------ = ----               P2       T2 The equation may be rearranged to determine P1:

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 T 1 P 1 = P 2 ------ T 2 For the NuScale design, P2 at 1,200 psia (design pressure), T1 at 100 degree F (559.7 degree R), and T2 representing a 0.1 °F degree change at 99.9 degree F (559.6 degree R) yields a pressure of: 559.7 °R P 1 = ( 1200 psia ) --------------------- = 1200.214 psia 559.6 °R The equivalent average temperature change in a large standard plant design with a 60 psia test pressure would have a resulting pressure of: 559.7 °R P 1 = ( 60 psia ) --------------------- = 60.011 psia 559.6 °R Thus, for the NuScale design, a 0.1 degree F undetected temperature increase in the gas volume would cause a corresponding pressure rise of 0.214 psi. The allowable pressure change to meet the leakage criteria for the NuScale design is approximately 0.06 psia. Therefore, a 0.1 degree F change in average temperature of the gas in the CNV results in more than three times the allowable pressure change associated with the maximum allowable leak rate at 1,200 psia. Changes in average temperature of the fluid inside containment have a similar although less pronounced impact. The impact of temperature emphasizes the need for highly accurate temperature measurements to obtain a representative average temperature of the CNV atmosphere during ILRT. It also highlights a challenge in obtaining accurate pressure measurement as high-precision gauges available for field installation are typically accurate within 0.01 percent of full scale, or 0.1 psia for a nominal 1,200 psia measurement. The pressure change of 0.011 psi for a large standard plant is more than 19 times smaller and would not be expected to cause failure of ILRT. D.4 Instrumentation Sensors to measure dew point temperature or relative humidity are not currently included in the CNV or RPV instrumentation. Multiple dew point sensors to perform the ILRT are needed in various regions and elevations, such as: near the top of the CNV where the insulated head is above the reactor pool surface mid-height of the gas volume just above the top of the internal CNV water level to ensure these different environments are monitored inside the PZR and possibly under the baffle plate to monitor the RPV gas space © Copyright 2022 by NuScale Power, LLC D-3

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 In general, ensuring proper placement, accuracy, and calibration is difficult. Furthermore, meeting ANSI/ANS 56.8 criteria is impractical and even doing so appears insufficient for providing accurate test results. Dew point sensors capable of withstanding 500 to 1,200 psia are uncommon but available. However, the most accurate sensors (e.g., gravimetric, chilled mirror hygrometer) are not suitable for field installation and usually are designed for applications involving flow past the sensor. Without dew point or relative humidity measurements, the effect of evaporative increases in vapor pressure would have to be approximated for the ILRT. Allowance for uncertainties due to these effects are included in establishing acceptance criteria. These allowances apply even if the plant conditions requiring them do not occur during the test (i.e., even if no temperature or dew point variations actually exist, the data would have to be adjusted on the assumption that the lack of ability to sense such variations was the result of insufficient monitoring capability). D.5 Leak Rate Criteria The allowable leak rates of large PWRs are typically above 1 standard cubic foot per minute (scfm) with many being around 5 scfm. The test acceptance criterion for NuScale is approximately 0.226 scfm at 1000 psia. This leak rate value is specified for illustrative purposes; it is based on the latest available references and is updated as the NPM-20 design matures. As a result, ILRT for an NPM must include monitoring accuracy that is 27 times better than commonly used. Because large PWRs sometimes have difficulty meeting their acceptance criteria for stability or accuracy, the challenge for NuScale is even greater. The acceptance criterion for passing ILRT is 0.75La. Typically, actual leakage is 30 to 50 percent of this acceptance criterion with the remainder reserved for operational margin to allow for some allowable degradation over future operating cycles without requiring immediate repair. Because the combination of uncertainty values may result in reducing the acceptable leak rate result to less than half of 0.75La, it is more likely that ILRT on NPMs will fail repeatedly on assumed and actual data uncertainty. Subsequent ILRT would need to be re-performed at a considerably higher rate than existing plants. Normal operational instrumentation provides insufficient accuracy and coverage for ILRT. More accurate sensors are needed for NPM integrated leak rate testing because the leak rate to be detected is approximately one-thirtieth of that for a large PWR. The sensors must also function at 1,000 psia. Such instrumentation must be assumed to be either permanently or temporarily installed inside the module rather than being located outside with sensors inserted through external test points, the inclusion of which would complicate the CNV design. © Copyright 2022 by NuScale Power, LLC D-4

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 D.6 Alternate Testing Arrangements Testing under dry containment conditions was also considered. Such conditions could be achieved through a complete core offload, or by completing the test with the ECCS valves closed and decay heat removal through the normal operating pathways (i.e., the SGs and, to a lesser extent, the CVCS). A full-core offload would eliminate impacts from core heat of the NPM being tested, but would not eliminate heat transfer from the reactor pool to the containment gas space. Numerous temperature and humidity measurements throughout the containment would still be required, which presents the same challenges already discussed. This approach also defeats the purpose of an as-found or as-left containment leakage test as neither cannot reasonably be truly performed if a full core offload is first required. Testing with the ECCS valves closed would require reactor pressure to be safely above the containment test pressure because of the passive nature of the ECCS valves, which begin to open when CNV pressure is close to or above RPV pressure. Reactor pressure, therefore, has to be greater than approximately 1,100 psia with a corresponding PZR temperature of approximately 556 degree F. This scenario presents an unacceptable negative safety impact to ECCS operation and introduces additional significant impacts to containment gas space temperature. Testing in these conditions also does not eliminate the challenges already presented. D.7 Conclusions NuScale reviewed the requirements of GDC 52 and 10 CFR 50, Appendix J, Type A testing to assess the potential of performing ILRT with the design. The inherent safety features of the NPM limit the ability of the design to conform with 10 CFR 50, Appendix J, Type A testing acceptance criteria and limit the effectiveness of Type A tests for the design. The heat transfer mechanisms and high heat transfer ability of the NPM creates a variable temperature and pressure atmosphere within containment. The prescriptive 10 CFR 50, Appendix J, Type A testing requirements and acceptance criteria are impractical for the design. The temperature and pressure impacts on Type A testing and associated acceptance criteria for the design increases the likelihood of inaccurate results, false test failures, and multiple iterations of testing. Application of Type A testing requirements to the CNV would likely yield inaccurate leakage results because of the limited effectiveness of Type A acceptance criteria when applied to the design. The evaluation of bolted flange connections provides reasonable assurance that the Type B measured leakage is representative of CNV leakage at design basis conditions. Additionally, the sealing studies performed on the CNV flange as part of technology maturation further justifies that the CNV design does not require Type A testing. Accessibility constraints within containment and the installation of a large quantity of additional CNV instrumentation (permanent or temporary) for Type A testing would expose occupational radiation workers to unnecessary radiation doses to support testing without a commensurate safety benefit. This unnecessary exposure would be required to support installation, maintenance, and calibration of the equipment necessary to perform © Copyright 2022 by NuScale Power, LLC D-5

NuScale Containment Leakage Integrity Assurance TR-123952-NP Revision 0 Type A tests. These dose impacts would multiply if additional ILRT is required following any failed tests. In summary, conformance with GDC 52 and 10 CFR 50, Appendix J, Type A testing requirements is impractical for the design. The Containment Leak Rate Testing Program, supported by the design, provides sufficient leakage integrity assurance for the containment. © Copyright 2022 by NuScale Power, LLC D-6}}

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