Text

Part 02 - Final Safety Analysis Report (Rev. 4) - Part 02 - Tier 02 - Chapter 03 - Design of Structures, Systems, Components and Equipment - Sections 03.01 - 03.06
	ML20036D423
Person / Time
Site:	NuScale
Issue date:	01/16/2020
From:	Bergman T; NuScale
To:	; Office of Nuclear Reactor Regulation
	Cranston G
References
	NUSCALESMRDC, NUSCALESMRDC.SUBMISSION.10, NUSCALEPART02.NP, NUSCALEPART02.NP.4
	Download: ML20036D423 (276)
	v • d • e

e Standard Plant Certification Application ter Three ign of Structures, tems, Components Equipment T 2 - TIER 2 4

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NRC is permitted to make the number of copies of the information contained in these reports ded for its internal use in connection with generic and plant-specific reviews and approvals, as well he issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or ation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding rictions on public disclosure to the extent such information has been identified as proprietary by cale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of e reports, the NRC is permitted to make the number of additional copies necessary to provide ies for public viewing in appropriate docket files in public document rooms in Washington, DC, and where as may be required by NRC regulations. Copies made by the NRC must include this copyright ce in all instances and the proprietary notice if the original was identified as proprietary.

NuScale Final Safety Analysis Report Table of Contents TABLE OF CONTENTS CHAPTER 3 DESIGN OF STRUCTURES, SYSTEMS, COMPONENTS AND EQUIPMENT. . .3.1-1 3.1 Conformance with U.S. Nuclear Regulatory Commission General Design Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-1 3.1.1 Overall Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-1 3.1.2 Protection by Multiple Fission Product Barriers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-7 3.1.3 Protection and Reactivity Control Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-16 3.1.4 Fluid Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-24 3.1.5 Reactor Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-37 3.1.6 Fuel and Radioactivity Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-43 3.1.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-47 3.2 Classification of Structures, Systems, and Components. . . . . . . . . . . . . . . . . . . . . . . . 3.2-1 3.2.1 Seismic Classification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-2 3.2.2 System Quality Group Classification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-4 3.2.3 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-7 3.3 Wind and Tornado Loadings. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1 3.3.1 Severe Wind Loadings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-1 3.3.2 Extreme Wind Loads (Tornado and Hurricane Loads). . . . . . . . . . . . . . . . . . . . . . . . . 3.3-2 3.3.3 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.3-4 3.4 Water Level (Flood) Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-1 3.4.1 Internal Flood Protection for Onsite Equipment Failures. . . . . . . . . . . . . . . . . . . . . . 3.4-1 3.4.2 Protection of Structures Against Flood from External Sources . . . . . . . . . . . . . . . . 3.4-7 3.4.3 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.4-9 3.5 Missile Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-1 3.5.1 Missile Selection and Description. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-1 3.5.2 Structures, Systems, and Components to be Protected from External Missiles. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-19 3.5.3 Barrier Design Procedures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-20 3.5.4 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-24 3.6 Protection against Dynamic Effects Associated with Postulated Rupture of Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-1 3.6.1 Plant Design for Protection against Postulated Piping Ruptures in Fluid Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-1 3.6.2 Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-8 Tier 2 i Revision 4

NuScale Final Safety Analysis Report Table of Contents TABLE OF CONTENTS 3.6.3 Leak-Before-Break Evaluation Procedures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-37 3.6.4 High Energy Line Break Evaluation (Non-LBB) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-61 3.6.5 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-62 3.7 Seismic Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-1 3.7.1 Seismic Design Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-2 3.7.2 Seismic System Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-105 3.7.3 Seismic Subsystem Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-444 3.7.4 Seismic Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-476 3.7.5 Computer Programs Used in Section 3.7 Seismic Design . . . . . . . . . . . . . . . . . . . 3.7-479 3.8 Design of Category I Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-1 3.8.1 Concrete Containment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-1 3.8.2 Steel Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-2 3.8.3 Concrete and Steel Internal Structures of Steel or Concrete Containments . . . 3.8-40 3.8.4 Other Seismic Category I Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-41 3.8.5 Foundations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-137 3.9 Mechanical Systems and Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-1 3.9.1 Special Topics for Mechanical Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-1 3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment . . . . 3.9-18 3.9.3 ASME Code Class 1, 2, and 3 Components, Component Supports, and Core Support Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-26 3.9.4 Control Rod Drive System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-35 3.9.5 Reactor Vessel Internals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-45 3.9.6 Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic Restraints. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-50 3.9.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-65 3.10 Seismic and Dynamic Qualifications of Mechanical and Electrical Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.10-1 3.10.1 Seismic Qualification Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.10-2 3.10.2 Methods and Procedures for Qualifying Mechanical and Electrical Equipment and Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.10-3 3.10.3 Methods and Procedures for Qualifying Supports of Mechanical and Electrical Equipment and Instrumentation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.10-6 3.10.4 Test and Analysis Results and Experience Database . . . . . . . . . . . . . . . . . . . . . . . . . 3.10-7 3.10.5 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.10-8 Tier 2 ii Revision 4

NuScale Final Safety Analysis Report Table of Contents TABLE OF CONTENTS 3.11 Environmental Qualification of Mechanical and Electrical Equipment . . . . . . . . . 3.11-1 3.11.1 Equipment Identification and Environmental Conditions. . . . . . . . . . . . . . . . . . . . 3.11-3 3.11.2 Qualification Tests and Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-5 3.11.3 Qualification Test Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-8 3.11.4 Loss of Ventilation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-8 3.11.5 Estimated Chemical and Radiation Environment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-9 3.11.6 Qualification of Mechanical Equipment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-10 3.11.7 Equipment Qualification Operational Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-11 3.11.8 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-12 3.12 ASME Code Class 1, 2, and 3 Piping Systems, Piping Components and Associated Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-1 3.12.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-1 3.12.2 Codes and Standards . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-2 3.12.3 Piping Analysis Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-3 3.12.4 Piping Modeling Technique . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-12 3.12.5 Piping Stress Analysis Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-16 3.12.6 Piping Support Design Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-25 3.12.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-30 3.13 Threaded Fasteners (ASME Code Class 1, 2, and 3) . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.13-1 3.13.1 Design Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.13-1 3.13.2 Inservice Inspection Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.13-3 3.13.3 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.13-4 Appendix 3A Dynamic Structural Analysis of the NuScale Power Module . . . . . . . . . . . . . . 3A-1 3A.1 Seismic Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3A-1 3A.2 Blowdown Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3A-1 Appendix 3B Design Reports and Critical Section Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-1 3B.1 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-4 3B.2 Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-16 3B.3 Control Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-32 3B.4 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-37 Appendix 3C Methodology for Environmental Qualification of Electrical and Mechanical Equipment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-1 3C.1 Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-1 Tier 2 iii Revision 4

NuScale Final Safety Analysis Report Table of Contents TABLE OF CONTENTS 3C.2 Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-1 3C.3 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-1 3C.4 Qualification Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-2 3C.5 Design Specifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-9 3C.6 Qualification Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-13 3C.7 Equipment Qualification Maintenance Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-16 3C.8 Documentation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-17 3C.9 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-17 Tier 2 iv Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3.2-1: Classification of Structures, Systems, and Components . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-8 Table 3.4-1: Flooding Sources in the Reactor Building and Control Building . . . . . . . . . . . . . . . . 3.4-10 Table 3.4-2: Flood Levels for Rooms Containing Systems, Structures, and Components Subject to Flood Protection (Without Mitigation) . . . . . . . . . . . . . . . . 3.4-11 Table 3.5-1: Concrete Thickness to Preclude Missile Penetration, Perforation, or Scabbing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-26 Table 3.5-2: Comparison of Test Case Missiles and NuScale Missiles . . . . . . . . . . . . . . . . . . . . . . . . 3.5-27 Table 3.5-3: Reactor Building Essential SSC Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-28 Table 3.5-4: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-30 Table 3.6-1: High- and Moderate-Energy Fluid System Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-64 Table 3.6-2: Postulated Break Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-67 Table 3.6-3a: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-68 Table 3.6-3b: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-69 Table 3.6-4: NuScale Power Module Piping Systems Design and Operating Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-70 Table 3.6-5: Mechanical Properties for Piping Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-71 Table 3.6-6: Allowable Stresses for Class 1 Piping (ksi) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-72 Table 3.6-7: Allowable Stresses for Class 2 & 3 Piping (ksi). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-73 Table 3.6-8: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-74 Table 3.7.1-1: Certified Seismic Design Response Spectra Control Points at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-19 Table 3.7.1-2: Certified Seismic Design Response Spectra - High Frequency Control Points at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-20 Table 3.7.1-3: Cross-Correlation Coefficients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-21 Table 3.7.1-4: Duration of Time Histories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-22 Table 3.7.1-5: Comparison of Response Spectra to CSDRS and CSDRS-HF . . . . . . . . . . . . . . . . . . . . 3.7-23 Table 3.7.1-6: Generic Damping Values for Dynamic Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-24 Table 3.7.1-7: Effective Stiffness of Reinforced Concrete Members . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-26 Table 3.7.1-7a: Effective Stiffness Changes of Cracked Reinforced Concrete Finite Element Model Members. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-27 Table 3.7.1-8: Soil Shear Modulus Degradation and Strain-Dependent Soil Damping (0-120 ft). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-28 Table 3.7.1-9: Soil Shear Modulus Degradation and Strain-Dependent Soil Damping (120 ft-1000 ft) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-29 Tier 2 v Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3.7.1-10: Strain-Dependent Soil Shear Moduli and Soil Damping Ratios for Gravel and Rock. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-30 Table 3.7.1-11: Soft Soil [Type 11] Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-31 Table 3.7.1-12: Firm Soil/Soft Rock [Type 8] Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-32 Table 3.7.1-13: Rock [Type 7] Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-33 Table 3.7.1-14: Hard Rock [Type 9] Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-34 Table 3.7.1-15: Average Strain-Compatible Properties for CSDRS for Rock [Type 7]. . . . . . . . . . . . . 3.7-35 Table 3.7.1-16: Average Strain-Compatible Properties for CSDRS for Soft Soil [Type 11] . . . . . . . . 3.7-37 Table 3.7.1-17: Average Strain-Compatible Properties for CSDRS for Firm Soil/Soft Rock

[Type 8]. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-39 Table 3.7.1-18: Strain-Compatible Properties for CSDRS-HF for Rock [Type 7] . . . . . . . . . . . . . . . . . . 3.7-41 Table 3.7.1-19: Strain-Compatible Properties for CSDRS-HF for Hard Rock [Type 9]. . . . . . . . . . . . . 3.7-43 Table 3.7.1-20: Wave Passing Frequencies. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-45 Table 3.7.1-21: Shear Wave Fundamental Frequencies of Soil Columns above RXB Foundation Bottom Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-46 Table 3.7.2-1: Summary of Reactor Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-148 Table 3.7.2-2: Average Hydrodynamic Pressure from ANSYS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-149 Table 3.7.2-3: Equivalent Average Static Pressure from SASSI2010 . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-150 Table 3.7.2-4: Summary of Average Pressures and Equivalent Static Pressure for SASSI2010 Soil Type 7. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-151 Table 3.7.2-5: Summary of Average Pressures and Equivalent Static Pressure for SASSI2010 Soil Type 8. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-152 Table 3.7.2-6: Summary of Average Pressures and Equivalent Static Pressure for SASSI2010 Soil Type 11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-153 Table 3.7.2-7: Comparison of Pressures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-154 Table 3.7.2-8: Final Surface Pressure Adjustment in SAP2000 Model Due to FSI Effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-155 Table 3.7.2-9: Summary of Control Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-156 Table 3.7.2-10: Summary of Reactor Building Fixed-Base Modal Frequency Comparison . . . . . . 3.7-157 Table 3.7.2-11: Summary of Control Building Fixed-Base Model Frequency Comparison . . . . . . 3.7-158 Table 3.7.2-12: Summary of Triple Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-159 Table 3.7.2-13: Dimensions and Weights of the Three Buildings. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-160 Table 3.7.2-14: Frequencies and Modal Mass Ratios for the Reactor Building Cracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-161 Tier 2 vi Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3.7.2-15: Frequencies and Modal Mass Ratios for the Reactor Building Uncracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-172 Table 3.7.2-16: Frequencies and Modal Mass Ratios for the Control Building Cracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-183 Table 3.7.2-17: Frequencies and Modal Mass Ratios for the Control Building Uncracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-188 Table 3.7.2-18: Frequencies Used in Transfer Function Calculation for Standalone Reactor Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-193 Table 3.7.2-19: Frequencies Used in Transfer Function Calculation for RXB from Triple Building Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-196 Table 3.7.2-20: Frequencies Used in Transfer Function Calculation for Standalone CRB Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-199 Table 3.7.2-21: Frequencies Used in Transfer Function Calculation for CRB with Triple Building CRB Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-203 Table 3.7.2-22: Methodology for Combining SASSI2010 Results. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-207 Table 3.7.2-23: Example Averaging and Bounding Forces and Moments in a Shell Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-209 Table 3.7.2-24: Example Averaging and Bounding Forces and Moments in a Beam Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-210 Table 3.7.2-25: Example Averaging and Bounding Forces and Moments in a Solid Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-212 Table 3.7.2-26: Selected Reactor Building Locations for Relative Displacement . . . . . . . . . . . . . . . 3.7-213 Table 3.7.2-27: Selected Control Building Locations for Relative Displacement Calculation. . . . 3.7-214 Table 3.7.2-28: Relative Displacement at Selected Locations on Reactor Building . . . . . . . . . . . . . 3.7-215 Table 3.7.2-29: Relative Displacement at Selected Locations on Control Building . . . . . . . . . . . . . 3.7-216 Table 3.7.2-30: Comparison of Maximum Lug and Skirt Reactions using Soil Type 7 (CSDRS). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-217 Table 3.7.2-31: Comparison of Maximum Lug and Skirt Reactions using Soil Type 9 (CSDRS-HF) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-218 Table 3.7.2-32: Max Forces and Moments at wall locations using Soil Type 7, CSDRS Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-219 Table 3.7.2-33: Definition of Seismic Analysis Identification Codes . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-220 Table 3.7.2-34: SSC Seismic Analysis Identification Code Assignments. . . . . . . . . . . . . . . . . . . . . . . . 3.7-221 Table 3.7.2-35: Analysis Model Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-222 Table 3.7.2-36: SASSI2010 3D Equivalent Stick Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-227 Table 3.7.2-37: ANSYS 3D Finite Element Beam Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-228 Tier 2 vii Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3.7.2-38: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-229 Table 3.7.2-39: Comparison of Lug Reactions due to Capitola Input for Model A and Model B. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-230 Table 3.7.2-40: Comparison of Maximum Out-of-Plane Shears and Moments due to Capitola Input in RXB Exterior Walls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-231 Table 3.7.2-41: Comparison of Maximum Out-of-Plane Shear Forces and Moments in CRB Exterior Walls due to Capitola Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-232 Table 3.7.2-42: Total Vertical Seismic RXB Base Reactions due to Capitola Input . . . . . . . . . . . . . . 3.7-233 Table 3.7.2-43: Total Vertical Seismic CRB Base Reactions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-234 Table 3.7.2-44: Relative Displacement at Critical Locations of Standalone CRB Model . . . . . . . . . 3.7-235 Table 3.7.2-45: Comparison of Maximum Forces and Moments in the North Pool Wall near the North Lug Support of RXM1 between 7P and DM . . . . . . . . . . . . . . . . . . . . 3.7-236 Table 3.7.2-46: Comparison of Maximum Seismic Out-of-Plane Shear Forces and Moments in CRB Exterior Walls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-237 Table 3.7.2-47: Comparison of RXB Relative Displacements between 7P and DM . . . . . . . . . . . . . 3.7-238 Table 3.7.2-48: North RXB Wall Soil Pressure Comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-239 Table 3.7.2-49: Building Models Used for RXB Basemat Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-240 Table 3.7.2-50: Basemat Model Used for RXB Basemat Design. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-241 Table 3.7.2-51: Building Models Used for CRB Basemat Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-242 Table 3.7.2-52: Basemat Model Used for CRB Basemat Design. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-243 Table 3.7.2-53: Floor Elevation and Nodes for Floor ISRS Generation . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-245 Table 3.7.2-54: SASSI Containment Vessel Skirt Coordinates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-246 Table 3.7.2-55: SASSI Containment Vessel Lug Coordinates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-247 Table 3.7.2-56: Selected Crane Wheel Locations and a Crane Rail Slab Node for In-Structure Response Spectra Presentation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-248 Table 3.7.2-57: Coordinates of Standalone and Triple Building Models for Control Building Floor In-Structure Response Spectra Generation. . . . . . . . . . . . . . . . . . . . . 3.7-249 Table 3.7.2-58: Coordinates of Selected Reactor Flange Tool Nodes . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-250 Table 3.7.2-59: Comparison of Empty Dry Dock Condition Lug Reactions with Final Safety Analysis Report Results (Capitola Input and Nominal NuScale Power Module Stiffness). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-251 Table 3.7.2-60: Comparison of Maximum Empty Dry Dock Condition Forces in NuScale Power Module Skirt Supports with Final Safety Analysis Report Results (Capitola Input and Nominal NuScale Power Module Stiffness). . . . . . . . . . . . . . . . 3.7-252 Table 3.7.3-1: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-453 Table 3.7.3-2: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-454 Tier 2 viii Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3.7.3-3: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-455 Table 3.7.3-4: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-456 Table 3.7.3-5: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-457 Table 3.7.3-6: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-458 Table 3.7.3-7: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-459 Table 3.7.3-8: Bioshield Nominal Dimensions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-460 Table 3.7.3-9: Bioshield Concrete and Reinforcement Design Properties . . . . . . . . . . . . . . . . . . . . 3.7-461 Table 3.7.3-10: Moment and Shear Capacity of Horizontal Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-462 Table 3.7.3-11: Bioshield Slab Self-Weight. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-463 Table 3.7.3-12: Bioshield Vertical Assembly Self-Weight for Structural Analysis . . . . . . . . . . . . . . . 3.7-464 Table 3.7.3-13: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-465 Table 3.7.3-14: Summary of Bioshield Demand to Capacity Ratios. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-466 Table 3.8.2-1: Design and Operating Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-26 Table 3.8.2-2: Load Combinations for Containment Vessel and Support ASME Code Stress Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-27 Table 3.8.2-3: Load Combinations for Containment Vessel Bolt ASME Code Stress Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-28 Table 3.8.2-4: Key Assumptions for CNV Ultimate Pressure Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-29 Table 3.8.4-1: Concrete Design Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-78 Table 3.8.4-2: Steel Design Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-79 Table 3.8.4-3: Summary of Reactor Building Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-80 Table 3.8.4-4: Summary of Control Building Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-81 Table 3.8.4-5: Hydrodynamic Weight . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-82 Table 3.8.4-6: Reactor Building SAP2000 Joints and Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-83 Table 3.8.4-7: Reactor Building SAP2000 Mass Sources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-84 Table 3.8.4-8: Control Building SAP2000 Joints and Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-85 Table 3.8.4-9: Control Building SAP2000 Mass Sources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-86 Table 3.8.4-10: Material Properties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-87 Table 3.8.4-11: Additional Dynamic Analyses. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-88 Table 3.8.4-12: Seismic Categories and Design Codes. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-89 Table 3.8.4-13: Total Weight in Kips, SAP2000 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-90 Table 3.8.4-14: Total Weight in Kips, SASSI2010 Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-91 Table 3.8.4-15: Seismic and Static Soil Pressures on CRB Walls for Standalone Model . . . . . . . . . . 3.8-92 Tier 2 ix Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3.8.4-16: Seismic and Static Soil Pressures on CRB Walls for Triple Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-93 Table 3.8.4-17: Enveloping Seismic Soil Pressures on CRB Walls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-94 Table 3.8.4-18: Seismic and Static Soil Pressures on RXB Walls for Standalone Model . . . . . . . . . . 3.8-95 Table 3.8.4-19: Seismic and Static Soil Pressures on RXB Walls for Triple Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-96 Table 3.8.4-20: Enveloping Seismic Soil Pressures on RXB Walls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-97 Table 3.8.4-21: RFT Embed Plates Demand to Capacity Ratios for SSE . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-98 Table 3.8.4-22: RFT Structural Member Demand to Capacity Ratios for SSE . . . . . . . . . . . . . . . . . . . . 3.8-99 Table 3.8.4-23: Reactor Flange Tool Structural Member Load Combinations . . . . . . . . . . . . . . . . . . 3.8-100 Table 3.8.5-1: RXB Stability Evaluation Input Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-160 Table 3.8.5-2: Reactor Building Static Effective Soil Force . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-161 Table 3.8.5-3: Seismic Base Reactions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-162 Table 3.8.5-4: Seismic Vertical RXB Base Reactions and Dead Weight. . . . . . . . . . . . . . . . . . . . . . . . 3.8-164 Table 3.8.5-5: Factors of safety - RXB Stability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-165 Table 3.8.5-6: RXB ANSYS Model Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-166 Table 3.8.5-7: Overturning Forces and Overturning Arms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-167 Table 3.8.5-7a: Displacement at Bottoms of Foundations of Uncracked Triple Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-168 Table 3.8.5-7b: Displacement at Bottoms of Foundations of Cracked Triple Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-169 Table 3.8.5-7c: Summary of Foundation Settlement of Uncracked Triple Building Model. . . . . . 3.8-170 Table 3.8.5-7d: Summary of Foundation Settlement of Cracked Triple Building Model . . . . . . . . 3.8-171 Table 3.8.5-8: CRB Stability Input Evaluation Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-172 Table 3.8.5-9: CRB Total Static Lateral Soil Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-173 Table 3.8.5-10: CRB SAP2000, SASSI2010, and ANSYS Model Summary . . . . . . . . . . . . . . . . . . . . . . . 3.8-174 Table 3.8.5-11: Reactor Building Sliding Displacements for Soil Type 7, 8 and 11 (Dead Weight + Buoyancy) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-175 Table 3.8.5-12: Control Building Sliding and Uplift Displacements for Soil Type 7 and 11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-176 Table 3.8.5-13: Average Soil Bearing Pressures (Toe Pressures) along Edges of RXB Basemat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-177 Table 3.8.5-14: Seismic Vertical CRB Base Reactions and Dead Weight. . . . . . . . . . . . . . . . . . . . . . . . 3.8-178 Table 3.8.5-15: Average Soil Bearing Pressures (Toe Pressures) along Edges of CRB Basemat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-179 Tier 2 x Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3.8.5-16: Reactor Building SAP2000 Basemat Model Summary . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-180 Table 3.8.5-17: CRB Tunnel Foundation Corner Displacements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-181 Table 3.8.5-18: CRB Tunnel Differential Settlement over 50 Feet and Tilt Angle . . . . . . . . . . . . . . . 3.8-182 Table 3.8.5-19: Foundation Sizes. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-183 Table 3.9-1: Summary of Design Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-66 Table 3.9-2: Pressure, Mechanical, and Thermal Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-67 Table 3.9-3: Required Load Combinations for Reactor Pressure Vessel American Society of Mechanical Engineers Stress Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-68 Table 3.9-4: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-69 Table 3.9-5: Required Load Combinations for Reactor Vessel Internals American Society of Mechanical Engineers Stress Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-70 Table 3.9-6: Required Load Combinations for Control Rod Drive Mechanism American Society of Mechanical Engineers Stress Analysis . . . . . . . . . . . . . . . . . . . . . 3.9-71 Table 3.9-7: Load Combinations for Decay Heat Removal System Condenser . . . . . . . . . . . . . . . 3.9-72 Table 3.9-8: Load Combinations for NuScale Power Module Top Support Structure. . . . . . . . . 3.9-73 Table 3.9-9: Loading Combinations for Decay Heat Removal System Actuation Valves . . . . . . 3.9-74 Table 3.9-10: Loads and Load Combinations for Reactor Safety Valves. . . . . . . . . . . . . . . . . . . . . . . 3.9-75 Table 3.9-11: Load Combinations for Emergency Core Cooling System Valves . . . . . . . . . . . . . . . 3.9-76 Table 3.9-12: Required Loads and Load Combinations for Secondary System Containment Isolation Valves. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-77 Table 3.9-13: Required American Society of Mechanical Engineers Code Loads and Load Combinations for Primary System Containment Isolation Valves. . . . . . . . . . 3.9-78 Table 3.9-14: Loads and Load Combinations for Thermal Relief Valves. . . . . . . . . . . . . . . . . . . . . . . 3.9-79 Table 3.9-15: Active Valve List. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-80 Table 3.9-16: Valve Inservice Test Requirements per ASME OM Code . . . . . . . . . . . . . . . . . . . . . . . . 3.9-82 Table 3.9-17: Valve Augmented Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-91 Table 3.9-18: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-94 Table 3.9-19: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-95 Table 3.9-20: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-96 Table 3.9-21: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-97 Table 3.9-22: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-98 Table 3.9-23: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-99 Table 3.9-24: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-100 Table 3.9-25: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-101 Tier 2 xi Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3.9-26: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-102 Table 3.11-1: List of Environmentally Qualified Electrical/I&C and Mechanical Equipment Located in Harsh Environments. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-14 Table 3.11-2: Environmental Qualification Zones - Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . 3.11-27 Table 3.12-1: Required Load Combinations for Class 1 Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-32 Table 3.12-2: Required Load Combinations for Class 2 & 3 Piping. . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-34 Table 3.12-3: Required Load Combinations for Class 1, 2, & 3 Supports . . . . . . . . . . . . . . . . . . . . . 3.12-35 Table 3.13-1: ASME BPV Code Section III Criteria for Selection and Testing of Bolted Materials. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.13-5 Table 3.13-2: ASME BPV Code Section XI Examination Categories for Inservice Inspections of Mechanical Joints in ASME Code Class 1, 2, and 3 Systems that are Secured by Threaded Fasteners. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.13-6 Table 3B-1: Identification of SAP2000 and SASSI2010 Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-38 Table 3B-2: Summary of D/C Ratios for Reactor Building Wall at Grid Line 1 . . . . . . . . . . . . . . . . .3B-39 Table 3B-2a: Magnitudes of Bounding Static, Dynamic, and Hydrodynamic Forces and Moments for RXB Wall at Grid Line 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-41 Table 3B-2b: Magnitudes of Bounding Final Design Forces and Moments for RXB Wall at Grid Line 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-43 Table 3B-3: Summary of D/C Ratios for Reactor Building Wall at Grid Line 3 . . . . . . . . . . . . . . . . .3B-44 Table 3B-3a: Magnitudes of Bounding Static, Dynamic, and Hydrodynamic Forces and Moments for RXB Wall at Grid Line 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-45 Table 3B-3b: Magnitudes of Bounding Final Design Forces and Moments for RXB Wall at Grid Line 3. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-46 Table 3B-4: Element Averaging of Horizontal Reinforcement Exceedance for Reactor Building Wall at Grid Line 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-47 Table 3B-5: Element Averaging of Horizontal Membrane Compression Stress for Reactor Building Wall at Grid Line 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-48 Table 3B-6: Element Averaging of Vertical Reinforcement Exceedance for Reactor Building Wall at Grid Line 3. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-49 Table 3B-7: Summary of D/C Ratios for Reactor Building Wall at Grid Line 3 After Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-50 Table 3B-8: Summary of D/C Ratios for Reactor Building Wall at Grid Line 4 . . . . . . . . . . . . . . . . .3B-51 Table 3B-8a: Magnitudes of Bounding Static, Dynamic, and Hydrodynamic Forces and Moments for RXB Wall at Grid Line 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-52 Table 3B-8b: Magnitudes of Bounding Final Design Forces and Moments for RXB Wall at Grid Line 4. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-53 Tier 2 xii Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3B-9: Element Averaging of Reinforcement Exceedance for Reactor Building Wall at Grid Line 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-54 Table 3B-10: Summary of D/C Ratios for RXB Wall at Grid Line 4 After Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-55 Table 3B-11: Summary of D/C Ratios for RXB Wall at Grid Line 6. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-56 Table 3B-11a: Magnitudes of Bounding, Dynamic, and Hydrodynamic Forces and Moments for RXB Wall at Grid Line 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-58 Table 3B-11b: Magnitudes of Bounding Final Design Forces and Moments for RXB Wall at Grid Line 6. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-60 Table 3B-12: Element Averaging of Horizontal Reinforcement Exceedance for RXB Wall at Grid Line 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-61 Table 3B-13: Summary of D/C Ratios for Reactor Building Wall at Grid Line 6 after Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-62 Table 3B-14: Summary of D/C Ratios for Reactor Building Wall at Grid Line E . . . . . . . . . . . . . . . . .3B-64 Table 3B-14a: Magnitudes of Bounding Static, Dynamic, and Hydrodynamic Forces and Moments for RXB Wall at Grid Line E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-67 Table 3B-14b: Magnitudes of Bounding Final Design Forces and Moments for RXB Wall at Grid Line E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-69 Table 3B-15: Summary of D/C Ratios for Reactor Building Slab at EL. 100-0. . . . . . . . . . . . . . . . . .3B-70 Table 3B-15a: Magnitudes of Bounding Static, Dynamic, and Hydrodynamic Forces and Moments for RXB Slab at EL. 100'-0". . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-72 Table 3B-15b: Magnitudes of Bounding Final Design Forces and Moments for RXB Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-73 Table 3B-16: Element Averaging of XZ Plane Shear Exceedance for Reactor Building Slab at EL. 100-0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-74 Table 3B-17: Summary of D/C Ratios for Reactor Building Slab at EL. 100-0 After Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-75 Table 3B-18: Summary of D/C Ratios for RXB Roof Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-77 Table 3B-18a: Magnitudes of Bounding Static, Dynamic, and Hydrodynamic Forces and Moments for RXB Roof Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-79 Table 3B-18b: Magnitudes of Bounding Final Design Forces and Moments for RXB Roof . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-80 Table 3B-19: Summary of D/C Ratios for Reactor Building Pilasters on Grid Line A Wall. . . . . . . .3B-81 Table 3B-19a: Magnitudes of Bounding Static, Dynamic, and Hydrodynamic Forces and Moments for RXB Pilasters on Grid Line A Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-83 Table 3B-19b: Magnitudes of Bounding Final Design Forces and Moments for RXB Pilasters on Grid Line A Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-86 Tier 2 xiii Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3B-20: Summary of D/C Ratios for Reactor Building Beams on EL. 75'-0" Slab . . . . . . . . . . .3B-87 Table 3B-20a: Magnitudes of Bounding Static, Dynamic, and Hydrodynamic Forces and Moments for RXB Beams on EL. 75'-0" Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-89 Table 3B-20b: Magnitudes of Bounding Final Design Forces and Moments for RXB Beams on EL. 75'-0" Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-90 Table 3B-21: Summary of D/C Ratios for Reactor Building Buttress at Grid Line 1 on EL. 126'-0" Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-91 Table 3B-21a: Magnitudes of Bounding Static, Dynamic, and Hydrodynamic Forces and Moments for RXB Buttress at Grid Line 1 on EL. 126'-0" Slab. . . . . . . . . . . . . . . . . . . . .3B-92 Table 3B-21b: Magnitudes of Bounding Final Design Forces and Moments for RXB Buttress at Grid Line 1 on EL. 126'-0" Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-93 Table 3B-22: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-94 Table 3B-22a: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-95 Table 3B-22b: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-96 Table 3B-23: Summary of D/C Ratios for Reactor Building Pool Wall at Grid Line B . . . . . . . . . . . .3B-97 Table 3B-23a: Magnitudes of Bounding Static, Dynamic, and Hydrodynamic Forces and Moments for RXB Pool Wall at Grid Line B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-99 Table 3B-23b: Magnitudes of Bounding Final Design Forces and Moments for RXB Pool Wall at Grid Line B. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-100 Table 3B-24: Element Averaging of YZ Plane Shear Exceedance for Reactor Building Pool Wall at Grid Line B. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-101 Table 3B-25: Summary of D/C Ratios for Reactor Building Pool Wall at Grid Line B After Averaging Affected Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-102 Table 3B-26: NuScale Power Module Lug Support Model Cut Section Forces and Moments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-104 Table 3B-27: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-105 Table 3B-28: Enveloped NPM Lug Support and Skirt Support Reaction Forces Using Soil Type 7 (CSDRS) and Design Capacities (x103 kips) . . . . . . . . . . . . . . . . . . 3B-106 Table 3B-29: Summary of D/C Ratios for Control Building Wall at Grid Line 3 . . . . . . . . . . . . . . . 3B-107 Table 3B-29a: Magnitudes of Bounding Static and Dynamic Forces and Moments for CRB Wall at Grid Line 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-108 Table 3B-29b: Magnitudes of Bounding Final Design Forces and Moments for CRB Wall at Grid Line 3. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-109 Table 3B-30: Summary of D/C Ratios for Control Building Wall at Grid Line 4 . . . . . . . . . . . . . . . 3B-110 Table 3B-30a: Magnitudes of Bounding Static and Dynamic Forces and Moments for CRB Wall at Grid Line 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-111 Tier 2 xiv Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3B-30b: Magnitudes of Bounding Final Design Forces and Moments for CRB Wall at Grid Line 4. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-112 Table 3B-31: Control Building Wall at Grid Line 4 - Shell Element 786 with added Shear Reinforcement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-113 Table 3B-32: Summary of D/C Ratios for Control Building Wall at Grid Line 4 After Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-114 Table 3B-33: Summary of D/C Ratios for Control Building Wall at Grid Line A . . . . . . . . . . . . . . . 3B-116 Table 3B-33a: Magnitudes of Bounding Static and Dynamic Forces and Moments for CRB Wall at Grid Line A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-117 Table 3B-33b: Magnitudes of Bounding Final Design Forces and Moments for CRB Wall at Grid Line A. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-118 Table 3B-34: Element Averaging of IP Shear Exceedance of Control Building Wall at Grid Line A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-119 Table 3B-35: Moment and Shear Capacity: 5 Foot Thick Control Building Basemat Foundation (Type 1). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-120 Table 3B-36: Moment and Shear Capacity: 5 Foot Thick Control Building Basemat Foundation (Type 2). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-121 Table 3B-37a: Magnitudes of Bounding Static and Dynamic Forces and Moments for Perimeter of CRB Basemat Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-122 Table 3B-37b: Magnitudes of Bounding Final Design Forces and Moments for Perimeter of Main Control Building Basemat Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-123 Table 3B-38a: Magnitudes of Bounding Static and Dynamic Forces and Moments for Interior of CRB Basemat Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-124 Table 3B-38b: Magnitudes of Bounding Final Design Forces and Moments for Interior of Main Control Building Basemat Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-125 Table 3B-39a: Magnitudes of Bounding Static and Dynamic Forces and Moments for Basemat of CRB Tunnel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-126 Table 3B-39b: Magnitudes of Bounding Final Design Forces and Moments for Control Building Basemat of Control Building Tunnel . . . . . . . . . . . . . . . . . . . . . 3B-127 Table 3B-40: Design Check Control Building Basemat Foundation of Perimeter of the Main Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-128 Table 3B-41: Design Check Control Building Basemat Foundation of Interior of the Main Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-129 Table 3B-42: Design Check for Control Building Basemat Foundation for the Control Building Tunnel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-130 Table 3B-43: Summary of D/C Ratios for Control Building Slab at EL. 100'-0" . . . . . . . . . . . . . . . . 3B-131 Table 3B-43a: Magnitudes of Bounding Static and Dynamic Forces and Moments for CRB Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-132 Tier 2 xv Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3B-43b: Magnitudes of Bounding Final Design Forces and Moments for CRB Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-133 Table 3B-44: Element Averaging of East-West Reinforcement Exceedance - Control Building Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-134 Table 3B-45: Element Averaging of XZ Plane Shear Exceedance - Control Building Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-135 Table 3B-46: Element Averaging of YZ Plane Shear Exceedance - Control Building Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-136 Table 3B-47: Summary of D/C Ratios for Control Building Slab at EL. 100'-0" After Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-137 Table 3B-48: Element Averaging of Shear Friction Exceedance for Control Building Slab at EL. 100-0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-138 Table 3B-49: Summary of D/C Ratios for Control Building Pilasters on Grid Line 1 Wall . . . . . . 3B-139 Table 3B-49a: Magnitudes of Bounding Static and Dynamic Forces and Moments for CRB Pilasters on Grid Line 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-140 Table 3B-49b: Magnitudes of Bounding Final Design Forces and Moments for CRB Pilasters on Grid Line 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-141 Table 3B-50: Summary of D/C Ratios for Control Building T-Beams on EL. 120'-0" Slab . . . . . . 3B-142 Table 3B-50a: Magnitudes of Bounding Static and Dynamic Forces and Moments for CRB T-Beams on EL. 120'-0" Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-143 Table 3B-50b: Magnitudes of Bounding Final Design Forces and Moments for CRB T-Beams on EL. 120'-0" Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-144 Table 3B-51: Element Averaging of IP Shear Exceedance of Reactor Building Wall at Grid Line 3. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-145 Table 3B-52: Element Averaging of Shear Friction Exceedance of Reactor Building Wall at Grid Line 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-146 Table 3B-53: Analysis Cases for NuScale Power Modules. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-147 Table 3B-54: Strength Reduction Factors for Reinforced Concrete Design . . . . . . . . . . . . . . . . . . 3B-148 Table 3B-55: RXB Critical Sections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-149 Table 3B-56: CRB Critical Sections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-150 Table 3B-57: D/C Ratios for Structural Components of the Lug Supports . . . . . . . . . . . . . . . . . . . 3B-151 Table 3B-58: ANSYS RXB Reinforcing Steel and Liner Steel Elastic Strain Summary for T0 and Ta+Pa . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-152 Table 3B-59: ANSYS RXB Strain Based Concrete Design Check for SDH Loads. . . . . . . . . . . . . . . 3B-153 Table 3B-60: ANSYS RXB Reinforcing Steel and Liner Steel Elastic Strain Summary for Load Combination 10 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-154 Tier 2 xvi Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3B-61: ANSYS RXB Reinforcing Steel and Liner Steel Elastic Strain Summary for Load Combination 13 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-155 Table 3B-62: Combined Maximum Values for RXB Basemat Forces and Moments. . . . . . . . . . . 3B-156 Table 3B-63: Magnitudes of Bounding Static and Dynamic RXB Basemat Forces and Moments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-157 Table 3B-64: Design Check for Reactor Building Basemat Foundation for Perimeter Region. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-158 Table 3B-65: Design Check for Reactor Building Basemat Foundation for Interior Region . . . 3B-159 Table 3B-66: Design Summary - Wall at Grid Line 1, EL 24'-75'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-160 Table 3B-67: Design Summary - Wall at Grid Line 1, EL 75'-100' . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-161 Table 3B-68: Design Summary - Wall at Grid Line 1, EL 100'-145'-6" . . . . . . . . . . . . . . . . . . . . . . . . 3B-162 Table 3B-69: Design Summary - Wall at Grid Line 1, EL 145'-6"-181' . . . . . . . . . . . . . . . . . . . . . . . . 3B-163 Table 3B-70: Design Summary - Interior Weir Wall at Grid Line 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-164 Table 3B-71: Design Summary - 4'-Thick Interior Upper Stiffener Wall at Grid Line 3 . . . . . . . . 3B-165 Table 3B-72: Design Summary - 5'-Thick Interior Wall at Grid Line 4 . . . . . . . . . . . . . . . . . . . . . . . . 3B-166 Table 3B-73: Design Summary - Reactor Building 4'-Thick Interior Wall at Grid Line 4 . . . . . . . 3B-167 Table 3B-74: Design Summary - 4'-Thick Pool Wall at Grid Line 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-168 Table 3B-75: Design Summary - Pool Wall at Grid Line 6 above EL 123' . . . . . . . . . . . . . . . . . . . . . 3B-169 Table 3B-76: Design Summary - Pool Wall at Grid Line 6 below EL 123' . . . . . . . . . . . . . . . . . . . . . 3B-170 Table 3B-77: Design Summary - Pool Wall at Grid Line 6 - 7'-6" Thick Section below EL 123'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-171 Table 3B-78: Design Summary - Exterior Wall at Grid Line E below EL 50' . . . . . . . . . . . . . . . . . . . 3B-172 Table 3B-79: Design Summary - Exterior Wall at Grid Line E between EL 50' and EL 100'. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-173 Table 3B-80: Design Summary - Exterior Wall at Grid Line above EL 100' . . . . . . . . . . . . . . . . . . . 3B-174 Table 3B-81: Design Summary - Reactor Building Basemat Perimeter . . . . . . . . . . . . . . . . . . . . . . 3B-175 Table 3B-82: Design Summary - Reactor Building Basemat Interior. . . . . . . . . . . . . . . . . . . . . . . . . 3B-176 Table 3B-83: Design Summary - Reactor Building Slab at EL 100' . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-177 Table 3B-84: Design Summary - Reactor Building Roof Slab at EL 181' . . . . . . . . . . . . . . . . . . . . . . 3B-178 Table 3B-85: Design Summary - West Wing Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-179 Table 3B-86: Design Summary - Pool Wall at Grid Line B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-180 Table 3B-87: Design Summary - Control Building Interior Wall at Grid Line 3 . . . . . . . . . . . . . . . 3B-181 Table 3B-88: Design Summary - Control Building Exterior Wall at Grid Line 4 . . . . . . . . . . . . . . . 3B-182 Table 3B-89: Design Summary - Control Building Exterior Wall at Grid Line A . . . . . . . . . . . . . . . 3B-183 Tier 2 xvii Revision 4

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 3B-90: Design Summary - Control Building Basemat Perimeter . . . . . . . . . . . . . . . . . . . . . . 3B-184 Table 3B-91: Design Summary - Control Building Basemat Interior . . . . . . . . . . . . . . . . . . . . . . . . . 3B-185 Table 3B-92: Design Summary - Control Building 2'-Thick Slab at EL 100' . . . . . . . . . . . . . . . . . . . 3B-186 Table 3B-93: Design Summary - Control Building 3'-Thick Slab at EL 100' . . . . . . . . . . . . . . . . . . . 3B-187 Table 3B-94: Design Summary - Control Building Tunnel Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-188 Table 3C-1: Environmental Qualification Zones - Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . . .3C-20 Table 3C-2: Designated Harsh Environment Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-21 Table 3C-3: Designated Mild Environment Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-22 Table 3C-4: Equipment Post-Accident Operating Times . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-23 Table 3C-5: EQ Program Margin Requirements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-25 Table 3C-6: Normal Operating Environmental Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-26 Table 3C-7: Design Basis Event Environmental Conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-27 Table 3C-8: Limiting Design Basis Accident EQ Radiation Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-29 Tier 2 xviii Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.5-1: Plan View of Partial NuScale Plant Showing Turbine Missile Trajectory . . . . . . . . 3.5-31 Figure 3.5-2: Section View of RXB and TGB Showing Turbine Missile Barriers . . . . . . . . . . . . . . . 3.5-32 Figure 3.5-3: Section View of CRB and TGB Showing Turbine Missile Barriers . . . . . . . . . . . . . . . 3.5-33 Figure 3.6-1: Flowchart of methodology for evaluation of line breaks. . . . . . . . . . . . . . . . . . . . . . 3.6-75 Figure 3.6-2: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-76 Figure 3.6-3: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-77 Figure 3.6-4: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-78 Figure 3.6-5: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-79 Figure 3.6-6: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-80 Figure 3.6-7: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-81 Figure 3.6-8: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-82 Figure 3.6-9: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-83 Figure 3.6-10: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-84 Figure 3.6-11: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-85 Figure 3.6-12: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-86 Figure 3.6-13: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-87 Figure 3.6-14: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-88 Figure 3.6-15: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-89 Figure 3.6-16: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-90 Figure 3.6-17: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-91 Figure 3.6-18: Flow Chart for Piping Leak-Before-Break Evaluation. . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-92 Figure 3.6-19: Illustration of Pipe with a Circumferential Through-Wall Crack . . . . . . . . . . . . . . . 3.6-93 Figure 3.6-20: Henry-Fauske's Model of Two-Phase Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-94 Figure 3.6-21: Local and Global Surface Roughness and Turns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-95 Figure 3.6-22: Crack Opening Displacement-Dependent Effective Crack Morphology . . . . . . . 3.6-96 Figure 3.6-23: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Straight Pipe Base Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-97 Figure 3.6-24: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Pipe-to-Pipe Weld . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-98 Figure 3.6-25: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Pipe-to-Safe-End Weld. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-99 Figure 3.6-26: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 12 Straight Pipe Base Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-100 Tier 2 xix Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.6-27: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 12 Pipe-to-Safe-End Weld . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-101 Figure 3.6-28: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Elbow Base Metal. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-102 Figure 3.6-29: Smooth Bounding Analysis Curve for Nominal Pipe Size 4 Feedwater System Line Base Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-103 Figure 3.6-30: Smooth Bounding Analysis Curve for Nominal Pipe Size 4 Feedwater System Line Welds. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-104 Figure 3.6-31: Smooth Bounding Analysis Curve for Nominal Pipe Size 5 Feedwater System Line Base Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-105 Figure 3.6-32: Smooth Bounding Analysis Curve for Nominal Pipe Size 5 Feedwater System Line Welds. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-106 Figure 3.6-33: Application of BTP 3-4 Break Location Guidance in the NPM bay and RXB . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-107 Figure 3.6-34: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-108 Figure 3.6-35: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-109 Figure 3.6-36: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-110 Figure 3.6-37: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-111 Figure 3.6-38: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-112 Figure 3.6-39: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-113 Figure 3.6-40: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-114 Figure 3.7.1-1: NuScale Horizontal CSDRS at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-47 Figure 3.7.1-2: NuScale Vertical CSDRS at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-48 Figure 3.7.1-3: NuScale Horizontal CSDRS-HF at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . 3.7-49 Figure 3.7.1-4: NuScale Vertical CSDRS-HF at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-50 Figure 3.7.1-5a: Original Time Histories for Yermo East-West . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-51 Figure 3.7.1-5b: CSDRS Compatible Time Histories for Yermo East-West . . . . . . . . . . . . . . . . . . . . . . 3.7-52 Figure 3.7.1-5c: Original Time Histories for Yermo North-South . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-53 Figure 3.7.1-5d: CSDRS Compatible Time Histories for Yermo North-South . . . . . . . . . . . . . . . . . . . 3.7-54 Figure 3.7.1-5e: Original Time Histories for Yermo Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-55 Figure 3.7.1-5f: CSDRS Compatible Time Histories for Yermo Vertical . . . . . . . . . . . . . . . . . . . . . . . . 3.7-56 Figure 3.7.1-6a: Original Time Histories for Capitola East-West . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-57 Figure 3.7.1-6b: CSDRS Compatible Time Histories for Capitola East-West . . . . . . . . . . . . . . . . . . . . 3.7-58 Figure 3.7.1-6c: Original Time Histories for Capitola North-South. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-59 Tier 2 xx Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.1-6d: CSDRS Compatible Time Histories for Capitola North-South . . . . . . . . . . . . . . . . . . 3.7-60 Figure 3.7.1-6e: Original Time Histories for Capitola Vertical. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-61 Figure 3.7.1-6f: CSDRS Compatible Time Histories for Capitola Vertical. . . . . . . . . . . . . . . . . . . . . . . 3.7-62 Figure 3.7.1-7a: Original Time Histories for Chi-Chi East-West . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-63 Figure 3.7.1-7b: CSDRS Compatible Time Histories for Chi-Chi East-West . . . . . . . . . . . . . . . . . . . . . 3.7-64 Figure 3.7.1-7c: Original Time Histories for Chi-Chi North-South. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-65 Figure 3.7.1-7d: CSDRS Compatible Time Histories for Chi-Chi North-South. . . . . . . . . . . . . . . . . . . 3.7-66 Figure 3.7.1-7e: Original Time Histories for Chi-Chi Vertical. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-67 Figure 3.7.1-7f: CSDRS Compatible Time Histories for Chi-Chi Vertical. . . . . . . . . . . . . . . . . . . . . . . . 3.7-68 Figure 3.7.1-8a: Original Time Histories for Izmit East-West. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-69 Figure 3.7.1-8b: CSDRS Compatible Time Histories for Izmit East-West . . . . . . . . . . . . . . . . . . . . . . . . 3.7-70 Figure 3.7.1-8c: Original Time Histories for Izmit North-South . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-71 Figure 3.7.1-8d: CSDRS Compatible Time Histories for Izmit North-South . . . . . . . . . . . . . . . . . . . . . 3.7-72 Figure 3.7.1-8e: Original Time Histories for Izmit Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-73 Figure 3.7.1-8f: CSDRS Compatible Time Histories for Izmit Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-74 Figure 3.7.1-9a: Original Time Histories for El Centro East-West. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-75 Figure 3.7.1-9b: CSDRS Compatible Time Histories for El Centro East-West. . . . . . . . . . . . . . . . . . . . 3.7-76 Figure 3.7.1-9c: Original Time Histories for El Centro North South . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-77 Figure 3.7.1-9d: CSDRS Compatible Time Histories for El Centro North-South . . . . . . . . . . . . . . . . . 3.7-78 Figure 3.7.1-9e: Original Time Histories for El Centro Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-79 Figure 3.7.1-9f: CSDRS Compatible Time Histories for El Centro Vertical . . . . . . . . . . . . . . . . . . . . . . 3.7-80 Figure 3.7.1-10a: Original Time Histories for Lucerne East-West. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-81 Figure 3.7.1-10b: CSDRS-HF Compatible Time Histories for Lucerne East-West . . . . . . . . . . . . . . . . . 3.7-82 Figure 3.7.1-10c: Original Time Histories for Lucerne North-South . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-83 Figure 3.7.1-10d: CSDRS-HF Compatible Time Histories for Lucerne North-South. . . . . . . . . . . . . . . 3.7-84 Figure 3.7.1-10e: Original Time Histories for Lucerne Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-85 Figure 3.7.1-10f: CSDRS-HF Compatible Time Histories for Lucerne Vertical. . . . . . . . . . . . . . . . . . . . 3.7-86 Figure 3.7.1-11: Normalized Arias Intensity Curve of North-South Component of Izmit Time History. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-87 Figure 3.7.1-12a: Average Response Spectra East-West . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-88 Figure 3.7.1-12b: Average Response Spectra North-South . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-89 Figure 3.7.1-12c: Average Response Spectra Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-90 Tier 2 xxi Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.1-13a: Power Spectral Density Curves CSDRS Compatible Time Histories . . . . . . . . . . . . 3.7-91 Figure 3.7.1-13b: Power Spectral Density Curves CSDRS-HF Compatible Time Histories. . . . . . . . . 3.7-92 Figure 3.7.1-14: Soil Shear Modulus Degradation Curves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-93 Figure 3.7.1-15: Strain Dependent Soil Damping Curves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-94 Figure 3.7.1-16: Shear Wave Velocities for All Soil Types. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-95 Figure 3.7.1-17: Layered Soil Model Used for NuScale Power Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-96 Figure 3.7.1-18: Density for All Soil Types . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-97 Figure 3.7.1-19: Average Strain Compatible Vs Profiles for CSDRS Compatible Inputs. . . . . . . . . . 3.7-98 Figure 3.7.1-20: Strain Compatible Vs Profiles for CSDRS-HF Compatible Input. . . . . . . . . . . . . . . . 3.7-99 Figure 3.7.1-21: Strain Compatible Damping for Soil Type 7 for CSDRS Compatible Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-100 Figure 3.7.1-22: Strain Compatible Damping for Soil Type 8 for CSDRS Compatible Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-101 Figure 3.7.1-23: Strain Compatible Damping for Soil Type 11 for CSDRS Compatible Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-102 Figure 3.7.1-24: Comparison of Average Strain Compatible Damping for CSDRS Compatible Inputs. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-103 Figure 3.7.1-25: Comparison of Strain Compatible Damping for CSDRS-HF Compatible Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-104 Figure 3.7.2-1: Control Building, Reactor Building, and Radioactive Waste Building in Soil (Looking Northeast) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-253 Figure 3.7.2-2: Section View of Control Building, Reactor Building, and Radioactive Waste Building in Soil (Looking Northeast) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-254 Figure 3.7.2-3: Global Origin of Building Models . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-255 Figure 3.7.2-4: Reactor Building Model Showing Global X, Y, and Z Axes at Origin . . . . . . . . . . 3.7-256 Figure 3.7.2-5: Location at Northeast Corner on Top of Basemat used for 7P versus 9P Comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-257 Figure 3.7.2-6: Location at NPM 1 East Wing Wall at Lug Support used for 7P versus 9P Comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-258 Figure 3.7.2-7: Location at Center of Roof Slab used for 7P versus 9P Comparison . . . . . . . . . . 3.7-259 Figure 3.7.2-8: 7P Versus 9P Comparison at Northeast Corner on Top of Basemat . . . . . . . . . . . 3.7-260 Figure 3.7.2-9: 7P Versus 9P Comparison at NPM 1 East Wing Wall at Lug Support . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-261 Figure 3.7.2-10: 7P Versus 9P Comparison at Center of Roof Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-262 Figure 3.7.2-11: Reactor Building in Ground (Looking Northeast) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-263 Tier 2 xxii Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.2-12: Quarter View of Reactor Building in Ground . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-264 Figure 3.7.2-13: Longitudinal View of Half of Reactor Building in Ground . . . . . . . . . . . . . . . . . . . . 3.7-265 Figure 3.7.2-14: Transverse View of Half of Reactor Building in Ground (Looking Northeast). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-266 Figure 3.7.2-15: Reactor Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-267 Figure 3.7.2-16: Reactor Building SASSI2010 Model without Hidden Lines . . . . . . . . . . . . . . . . . . . 3.7-268 Figure 3.7.2-17: Reactor Building SASSI2010 Backfill Soil Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-269 Figure 3.7.2-18: Reactor Building SASSI2010 Model without Backfill . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-270 Figure 3.7.2-19: Reactor Building SASSI2010 Excavated Soil Model without Hidden Lines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-271 Figure 3.7.2-20: Half of Reactor Building SASSI2010 Model without Hidden Lines . . . . . . . . . . . . 3.7-272 Figure 3.7.2-21: Reactor Building Beam Elements of SASSI2010 Model. . . . . . . . . . . . . . . . . . . . . . . 3.7-273 Figure 3.7.2-22: NuScale Power Module Lug Restraint (in Green) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-274 Figure 3.7.2-23: Top View of NuScale Power Module Lug Restraint and Support Walls. . . . . . . . 3.7-275 Figure 3.7.2-24: View of Reactor Building Looking Down . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-276 Figure 3.7.2-25: Enlarged View of Reactor Pool Looking Down . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-277 Figure 3.7.2-26: Extruded View of the NuScale Power Modules and Support Walls . . . . . . . . . . . 3.7-278 Figure 3.7.2-27: NuScale Power Module Model with Lug Restraint and Base Skirt Supports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-279 Figure 3.7.2-28: NuScale Power Module Beam Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-280 Figure 3.7.2-29: Beam and Spring Model of Reactor Building Crane. . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-281 Figure 3.7.2-30: Longitudinal Section View of Pool Water and NuScale Power Modules . . . . . . 3.7-282 Figure 3.7.2-31: Model of Reactor Building Pool Water . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-283 Figure 3.7.2-32: Half Sectional View of Reactor Building ANSYS Model with Pool Fluid and Backfill Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-284 Figure 3.7.2-33: ANSYS Model of Fluid, NuScale Power Modules, Foundation and Interior Pool Walls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-285 Figure 3.7.2-34: 3D View of Pool without Water with 12 NuScale Power Modules . . . . . . . . . . . . 3.7-286 Figure 3.7.2-35: Plan View of Wall Segments used for FSI analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-287 Figure 3.7.2-36: Maximum Accelerations for X Wall Sections from ANSYS . . . . . . . . . . . . . . . . . . . . 3.7-288 Figure 3.7.2-37: Maximum Accelerations for Y Wall Sections from ANSYS . . . . . . . . . . . . . . . . . . . . 3.7-289 Figure 3.7.2-38: Hydrodynamic Pressure for X Wall Sections from ANSYS . . . . . . . . . . . . . . . . . . . . 3.7-290 Figure 3.7.2-39: Hydrodynamic Pressure for Y Wall Sections from ANSYS . . . . . . . . . . . . . . . . . . . . 3.7-291 Tier 2 xxiii Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.2-40: Maximum Accelerations for X Wall Sections, Soil Type 7 from SASSI2010. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-292 Figure 3.7.2-41: Maximum Accelerations for Y Wall Sections, Soil Type 7 from SASSI2010. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-293 Figure 3.7.2-42: Maximum Accelerations for X Wall Sections, Soil Type 8 from SASSI2010. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-294 Figure 3.7.2-43: Maximum Accelerations for Y Wall Sections, Soil Type 8 from SASSI2010. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-295 Figure 3.7.2-44: Maximum Accelerations for X Wall Sections, Soil Type 11 from SASSI2010. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-296 Figure 3.7.2-45: Maximum Accelerations for Y Wall Sections, Soil Type 11 from SASSI2010. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-297 Figure 3.7.2-46: Control Building (Looking Northeast). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-298 Figure 3.7.2-47: East-West Section Cut View of Control Building in Soil . . . . . . . . . . . . . . . . . . . . . . 3.7-299 Figure 3.7.2-48: North-South Section Cut View of Control Building in Soil. . . . . . . . . . . . . . . . . . . . 3.7-300 Figure 3.7.2-49: Quarter View of Control Building in Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-301 Figure 3.7.2-50: SAP2000 Control Building Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-302 Figure 3.7.2-51: SAP2000 Control Building Model Beam Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-303 Figure 3.7.2-52: SAP2000 Control Building Model with Backfill Soil . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-304 Figure 3.7.2-53: Control Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-305 Figure 3.7.2-54: Excavated Soil of Control Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . 3.7-306 Figure 3.7.2-55: Control Building SASSI2010 Model Backfill Soil Solid Elements . . . . . . . . . . . . . . 3.7-307 Figure 3.7.2-56: Control Building SASSI2010 Solid Elements Modeling the Basemat . . . . . . . . . . 3.7-308 Figure 3.7.2-57: Control Building SASSI2010 Model Shell and Beam Element . . . . . . . . . . . . . . . . 3.7-309 Figure 3.7.2-58: Control Building SASSI2010 Model Beam Elements . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-310 Figure 3.7.2-59: Structures and Backfill Soil of Triple Building SAP2000 Model . . . . . . . . . . . . . . . 3.7-311 Figure 3.7.2-60: Isometric View of South Side of Triple SAP2000 Model (Looking Northeast). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-312 Figure 3.7.2-61: Isometric View of South Side of Triple Building SAP2000 Model (Looking Northwest). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-313 Figure 3.7.2-62: Isometric View of North Side of Triple Building SAP2000 Model (Looking Southwest). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-314 Figure 3.7.2-63: Isometric View of Backfill Soil Elements around the Three Buildings . . . . . . . . . 3.7-315 Figure 3.7.2-64: Beam Elements of Triple Building SAP2000 Model . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-316 Figure 3.7.2-65: Spring or Link Elements of Triple Building SAP2000 Model . . . . . . . . . . . . . . . . . . 3.7-317 Tier 2 xxiv Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.2-66: Elevation View of Triple Building SAP2000 Model Showing Separation . . . . . . 3.7-318 Figure 3.7.2-67: Isometric View of SASSI2010 Triple Building . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-319 Figure 3.7.2-68: North Half View of SASSI2010 Triple Building Model . . . . . . . . . . . . . . . . . . . . . . . . 3.7-320 Figure 3.7.2-69: SASSI2010 Triple Building Model Shown without Backfill. . . . . . . . . . . . . . . . . . . . 3.7-321 Figure 3.7.2-70: SASSI2010 Triple Building Model Showing South Side of Three Buildings (Looking Northwest) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-322 Figure 3.7.2-71: SASSI2010 Triple Building Model Showing North Side of Three Buildings (Looking Southwest) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-323 Figure 3.7.2-72: Beam Elements of SASSI2010 Triple Building Model. . . . . . . . . . . . . . . . . . . . . . . . . 3.7-324 Figure 3.7.2-73: Excavated Soil Solid Elements of the SASSI2010 Triple Building Model. . . . . . . 3.7-325 Figure 3.7.2-74: Backfill Soil Solid Elements of the SASSI2010 Triple Building Model. . . . . . . . . . 3.7-326 Figure 3.7.2-75: Rigid Soil Springs between Backfill and Free Field Soils of the SASSI2010 Triple Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-327 Figure 3.7.2-76: Interaction Nodes for Soil Impedance Calculation. . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-328 Figure 3.7.2-77: Reactor Building Cracked Model - Modal Shape - (Mode 5) . . . . . . . . . . . . . . . . . . 3.7-329 Figure 3.7.2-78: Reactor Building Cracked Model - Modal Shape - (Mode 6) . . . . . . . . . . . . . . . . . . 3.7-330 Figure 3.7.2-79: Reactor Building Cracked Model - Modal Shape - (Mode 14) . . . . . . . . . . . . . . . . . 3.7-331 Figure 3.7.2-80: Reactor Building Uncracked Model - Modal Shape - (Mode 5) . . . . . . . . . . . . . . . 3.7-332 Figure 3.7.2-81: Reactor Building Uncracked Model - Modal Shape - (Mode 8) . . . . . . . . . . . . . . . 3.7-333 Figure 3.7.2-82: Reactor Building Uncracked Model - Modal Shape - (Mode 15) . . . . . . . . . . . . . . 3.7-334 Figure 3.7.2-83: Control Building Cracked Model - Modal Shape - (Mode 40) . . . . . . . . . . . . . . . . . 3.7-335 Figure 3.7.2-84: Control Building Cracked Model - Modal Shape - (Mode 49) . . . . . . . . . . . . . . . . . 3.7-336 Figure 3.7.2-85: Control Building Cracked Model - Modal Shape - (Mode 81) . . . . . . . . . . . . . . . . . 3.7-337 Figure 3.7.2-86: Control Building Uncracked Model - Modal Shape - (Mode 41) . . . . . . . . . . . . . . 3.7-338 Figure 3.7.2-87: Control Building Uncracked Model - Modal Shape - (Mode 49) . . . . . . . . . . . . . . 3.7-339 Figure 3.7.2-88: Control Building Uncracked Model - Modal Shape - (Mode 81) . . . . . . . . . . . . . . 3.7-340 Figure 3.7.2-89: Flow of Files among SASSI2010 Modules . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-341 Figure 3.7.2-90: Four I, J, K, and L Nodes and 1 through 5 Output Locations of Thick Shell Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-342 Figure 3.7.2-91: SASSI2010 Shear and Moment Resultants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-343 Figure 3.7.2-92: SASSI2010 Beam Element Local Axes for Forces and Moments . . . . . . . . . . . . . . 3.7-344 Figure 3.7.2-93: SASSI2010 Global Stresses at Centroid of a Solid Element . . . . . . . . . . . . . . . . . . . 3.7-345 Figure 3.7.2-94: Reactor Building Locations Selected for Relative Displacement. . . . . . . . . . . . . . 3.7-346 Tier 2 xxv Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.2-95: Control Building Locations Selected for Relative Displacement . . . . . . . . . . . . . . 3.7-347 Figure 3.7.2-96: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-348 Figure 3.7.2-97: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-349 Figure 3.7.2-98: Location of NPMs for 7 Module Case Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-350 Figure 3.7.2-99: Example ISRS from CSDRS compatible Time Histories for Soil Type 7 . . . . . . . . 3.7-351 Figure 3.7.2-100: Example ISRS from CSDRS compatible Time Histories for Soil Type 8 . . . . . . . . 3.7-352 Figure 3.7.2-101: Example ISRS from CSDRS compatible Time Histories for Soil Type 11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-353 Figure 3.7.2-102: Example Combined and Enveloped ISRS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-354 Figure 3.7.2-103: Exampled Broadened ISRS at Multiple Damping Ratios . . . . . . . . . . . . . . . . . . . . . 3.7-355 Figure 3.7.2-104: Comparison between Standalone Reactor Building Model and Triple Building Model, ISRS at Northwest Corner, Top of Basement . . . . . . . . . . . . . . . . 3.7-356 Figure 3.7.2-105: Comparison between Standalone Reactor Building Model and Triple Building Model, ISRS at Northwest Corner, Top of Exterior Wall. . . . . . . . . . . . . . 3.7-357 Figure 3.7.2-106: Comparison between Standalone Reactor Building Model and Triple Building Model, ISRS at Corner or Roof Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-358 Figure 3.7.2-107: Reactor Building ISRS for Floor at El. 24 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-359 Figure 3.7.2-108: Reactor Building ISRS for Floor at El. 25 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-360 Figure 3.7.2-109: Reactor Building ISRS for Floor at El. 50 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-361 Figure 3.7.2-110: Reactor Building ISRS for Floor at El. 75 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-362 Figure 3.7.2-111: Reactor Building ISRS for Floor at El. 100 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-363 Figure 3.7.2-112: Reactor Building ISRS for Floor at El. 126 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-364 Figure 3.7.2-113: Reactor Building ISRS for Roof at El. 181 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-365 Figure 3.7.2-114: In-Structure Response Spectra at Reactor Building Crane Wheels at El. 145 6. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-366 Figure 3.7.2-115: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-367 Figure 3.7.2-116: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-368 Figure 3.7.2-117: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-369 Figure 3.7.2-117a: CRB - ISRS at El. 50-0 (Z=405), 5 CSDRS Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-370 Figure 3.7.2-117b: CRB - ISRS at El. 50-0 (Z=405), CSDRS-HF Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-371 Figure 3.7.2-118: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-372 Figure 3.7.2-118a: Control Building - In-Structure Response Spectra at El. 63-3 (Z=570),

5 CSDRS Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-373 Tier 2 xxvi Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.2-118b: Control Building - In-Structure Response Spectra at El. 63-3 (Z=570),

CSDRS-HF Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-374 Figure 3.7.2-119: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-375 Figure 3.7.2-119a: Control Building - In-Structure Response Spectra at El. 76-6 (Z=720),

5 CSDRS Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-376 Figure 3.7.2-119b: CRB - East - West (X) ISRS at El. 76-6 (Z=720), CSDRS-HF Input . . . . . . . . . . . . . 3.7-377 Figure 3.7.2-120: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-378 Figure 3.7.2-120a: CRB - ISRS at El. 100-0 (Z=1020), 5 CSDRS Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-379 Figure 3.7.2-120b: CRB - ISRS at El. 100-0 (Z=1020), CSDRS-HF Input . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-380 Figure 3.7.2-121: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-381 Figure 3.7.2-121a: CRB - ISRS at El. 120-0 (Z=1260), 5 CSDRS Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-382 Figure 3.7.2-121b: CRB - ISRS at El. 120-0 (Z=1260), CSDRS-HF Input . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-383 Figure 3.7.2-122: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-384 Figure 3.7.2-122a: CRB - ISRS at El. 140-0 (Z=1518), 5 CSDRS Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-385 Figure 3.7.2-122b: CRB - ISRS at El. 140-0 (Z=1518), CSDRS-HF Input . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-386 Figure 3.7.2-123: Comparison of 12 NPM and 7 NPM Model Results at Northwest Corner on Top of Basement (EL. 24-0) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-387 Figure 3.7.2-124: Comparison of 12 NPM and 7 NPM Model Results at Mid-Span of North Wall on Top of Basement (EL. 24-0) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-388 Figure 3.7.2-125: Comparison of 12 NPM and 7 NPM Model Results at Northeast Corner on Top of Basement (EL. 24-0) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-389 Figure 3.7.2-126: Comparison of 12 NPM and 7 NPM Model Results at Northwest Corner on Top of Roof Slab (EL. 181-0) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-390 Figure 3.7.2-127: Comparison of 12 NPM and 7 NPM Model Results at Mid-Span of Roof Slab (EL. 181-0). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-391 Figure 3.7.2-128: Comparison of 12 NPM and 7 NPM Model Results at Northwest Corner of Roof Slab (EL. 181-0) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-392 Figure 3.7.2-129: Development of Average Static Pressure of 4.20 psi at Mid Height of Pool Water for SAP2000 Model to Account for 3D FSI Effects. . . . . . . . . . . . . . 3.7-393 Figure 3.7.2-130: RXB Node 3996, NW Corner on Top of Basemat, X-ISRS Comparison, Cracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-394 Figure 3.7.2-131: RXB Node 3996, NW Corner on Top of Basemat, Y-ISRS Comparison, Cracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-395 Figure 3.7.2-132: RXB Node 3996, NW Corner on Top of Basemat, Z-ISRS Comparison, Cracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-396 Tier 2 xxvii Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.2-133: RXB Node 3996, NW Corner on Top of Basemat, X-TF Comparison, Cracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-397 Figure 3.7.2-134: RXB Node 3996, NW Corner on Top of Basemat, Y-TF Comparison, Cracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-398 Figure 3.7.2-135: RXB Node 3996, NW Corner on Top of Basemat, Z-TF Comparison, Cracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-399 Figure 3.7.2-136: CRB - East-West (X) ISRS, Node 34380, Slab between Grid Line CB-D and CB-E at El. 63', Capitola Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-400 Figure 3.7.2-137: CRB - North-South (Y) ISRS, Node 34380, Slab between Grid Line CB-D and CB-E at El. 63', Capitola Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-401 Figure 3.7.2-138: CRB - Vertical (Z) ISRS, Node 34380, Slab between Grid Line CB-D and CB-E at El. 63', Capitola Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-402 Figure 3.7.2-139: Cracked CRB Transfer Function Amplitudes, X Response at Node 34380, Slab Between Grid Line CB-D and CB-E at El. 63' for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-403 Figure 3.7.2-140: Cracked CRB Transfer Function Amplitudes, Y Response at Node 34380, Slab Between Grid Line CB-D and CB-E at El. 63 for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-404 Figure 3.7.2-141: Cracked CRB Transfer Function Amplitudes, Z Response at Node 34380, Slab Between Grid Line CB-D and CB-E at El. 63 for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-405 Figure 3.7.2-142: Floor ISRS Locations at TOC EL 24'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-406 Figure 3.7.2-143: Floor Locations at TOC EL 25-0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-407 Figure 3.7.2-144: Floor ISRS Locations at TOC EL 50'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-408 Figure 3.7.2-145: Floor ISRS Locations at TOC EL 75' - 0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-409 Figure 3.7.2-146: Floor ISRS Locations at TOC EL 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-410 Figure 3.7.2-147: Floor ISRS Locations at TOC EL 126'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-411 Figure 3.7.2-148: Roof ISRS Locations at TOC EL 181'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-412 Figure 3.7.2-149: Non-Vertically Propagating Seismic Wave RXB Sensitivity Analysis Cases . . . . 3.7-413 Figure 3.7.2-150: Non-Vertically Propagating Seismic Wave Sensitivity Study with Soil Type 7 - Free Field Uncoupled East-West (X) ARS at Surface, Capitola Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-414 Figure 3.7.2-151: Non-Vertically Propagating Seismic Wave Sensitivity Study with Soil Type 7 - Free Field East-West (X) ARS Depth 85', Capitola Input, Alpha = 30 degrees . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-415 Tier 2 xxviii Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.2-152: Non-Vertically Propagating Seismic Wave Sensitivity Study with Soil Type 7 - Comparison of Combined Free-Field East-West (X) ARS at Depth 85', Capitola Input, Alpha = 0 Degrees, 17 Degrees, 30 Degrees . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-416 Figure 3.7.2-153: Non-Vertically Propagating Seismic Wave Sensitivity Study for the RXB - Uncombined East-West (X) ISRS, Node 3996, Top of Basemat, NW Corner, Capitola Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-417 Figure 3.7.2-154: Non-Vertically Propagating Seismic Wave Sensitivity Study for the RXB - Uncombined North-South (Y) ISRS, Node 3996, Top of Basemat, NW Corner, Capitola Input. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-418 Figure 3.7.2-155: Non-Vertically Propagating Seismic Wave Sensitivity Study for the RXB - Uncombined Vertical (Z) ISRS, Node 3996, Top of Basemat, NW Corner, Capitola Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-419 Figure 3.7.2-156: In-Structure Response Spectra at the Containment Vessel Skirt of NuScale Power Module 1, Capitola Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-420 Figure 3.7.2-157: In-Structure Response Spectra at the Containment Vessel Skirt of NuScale Power Module 6, Capitola Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-421 Figure 3.7.2-158: In-Structure Response Spectra at the East Lug of NuScale Power Module 1, Capitola Input. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-422 Figure 3.7.2-159: In-Structure Response Spectra at the North Lug of NuScale Power Module 1, Capitola Input. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-423 Figure 3.7.2-160: In-Structure Response Spectra at the West Lug of NuScale Power Module 1, Capitola Input. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-424 Figure 3.7.2-161: In-Structure Response Spectra at the East Lug of NuScale Power Module 6, Capitola Input. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-425 Figure 3.7.2-162: In-Structure Response Spectra at the North Lug of NuScale Power Module 6, Capitola Input. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-426 Figure 3.7.2-163: In-Structure Response Spectra at the West Lug of NuScale Power Module 6, Capitola Input. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-427 Figure 3.7.2-164: In-Structure Response Spectra at the Reactor Flange Tool Base, Node 6328, due to X, Y, and Z Inputs of Capitola Excitation for Soil Type 7, Uncracked Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-428 Figure 3.7.2-165: In-Structure Response Spectra at the Reactor Flange Tool Base, Node 6329, due to X, Y, and Z inputs of Capitola Excitation for Soil Type 7, Uncracked Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-429 Figure 3.7.2-166: In-Structure Response Spectra at the Reactor Flange Tool Base, Node 6330, due to X, Y, and Z Inputs of Capitola Excitation for Soil Type 7, Uncracked Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-430 Tier 2 xxix Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.2-167: In-Structure Response Spectra at the Reactor Flange Tool Base, Node 6331, due to X, Y, and Z Inputs of Capitola Excitation for Soil Type 7, Uncracked Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-431 Figure 3.7.2-168: In-Structure Response Spectra at the Reactor Flange Tool Base, Node 6328, due to X, Y, and Z Inputs of Capitola Excitation for Soil Type 7, Cracked Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-432 Figure 3.7.2-169: In-Structure Response Spectra at the Reactor Flange Tool Base, Node 6329, due to X, Y, and Z Inputs of Capitola Excitation for Soil Type 7, Cracked Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-433 Figure 3.7.2-170: In-Structure Response Spectra at the Reactor Flange Tool Base, Node 6330, due to X, Y, and Z Inputs of Capitola Excitation for Soil Type 7, Cracked Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-434 Figure 3.7.2-171: In-Structure Response Spectra at the Reactor Flange Tool Base, Node 6331, due to X, Y, and Z Inputs of Capitola Excitation for Soil Type 7, Cracked Condition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-435 Figure 3.7.2-172: Floor In-Structure Response Spectra at El. 50 ft, (Z=420 in.)

Comparing Full and Empty Dry Dock Conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-436 Figure 3.7.2-173: Floor In-Structure Response Spectra at El. 75 ft, (Z=720 in.)

Comparing Full and Empty Dry Dock Conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-437 Figure 3.7.2-174: Floor In-Structure Response Spectra at El. 100 ft, (Z=1020 in.)

Comparing Full and Empty Dry Dock Conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-438 Figure 3.7.2-175: In-Structure Response Spectra at Reactor Building Crane Wheel Node 29545 Comparing Full and Empty Dry Dock Conditions . . . . . . . . . . . . . . . 3.7-439 Figure 3.7.2-176a: In-Structure Response Spectra at the Northwest Corner of the Bioshield . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-440 Figure 3.7.2-176b: In-Structure Response Spectra at the Northeast Corner of the Bioshield . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-441 Figure 3.7.2-176c: In-Structure Response Spectra at the Northwest Corner of the Bioshield - High Frequency . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-442 Figure 3.7.2-176d: In-Structure Response Spectra at the Northeast Corner of the Bioshield - High Frequency . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-443 Figure 3.7.3-1: Cutoff View of Reactor Building to Depict Bioshields . . . . . . . . . . . . . . . . . . . . . . . . 3.7-467 Figure 3.7.3-2: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-468 Figure 3.7.3-2a: Bioshield Conceptual Design (Isometric View) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-469 Figure 3.7.3-2b: Bioshield Conceptual Design (Elevation View During Stacked Configuration). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-470 Figure 3.7.3-3: Location In-structure Response Spectra Nodes for Design of Bioshields. . . . . . 3.7-471 Figure 3.7.3-4: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-472 Tier 2 xxx Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.7.3-4a: In-Structure Response Spectra at the Bioshield in X-Direction for Nodes 25826 and 26345 and the Enveloped ISRS using 4% Damping . . . . . . . . 3.7-473 Figure 3.7.3-4b: In-Structure Response Spectra at the Bioshield in Y-Direction for Nodes 25826 and 26345 and the Enveloped ISRS using 4% Damping . . . . . . . . 3.7-474 Figure 3.7.3-4c: In-Structure Response Spectra at the Bioshield in Z-Direction for Nodes 25826 and 26345 and the Enveloped ISRS using 4% Damping . . . . . . . . 3.7-475 Figure 3.8.2-1: Containment Vessel Components and Building Elevations . . . . . . . . . . . . . . . . . . . 3.8-30 Figure 3.8.2-2: Passive Skirt Support Ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-31 Figure 3.8.2-3: Containment Vessel Lateral Lug Located within the NuScale Power Module Lug Restraints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-32 Figure 3.8.2-4: Containment Vessel Top Head Mechanical Penetrations . . . . . . . . . . . . . . . . . . . . . 3.8-33 Figure 3.8.2-5: Containment Vessel Top Head Instrumentation and Controls, Electrical, and Access Penetrations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-34 Figure 3.8.2-6: Typical Access Cover and O-Ring Seals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-35 Figure 3.8.2-7: Typical Non Secondary Side Containment Vessel Penetration Configuration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-36 Figure 3.8.2-8: Containment Vessel Reactor Pressure Vessel Support Boundary . . . . . . . . . . . . . . 3.8-37 Figure 3.8.2-9: Containment Vessel Bottom Head Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-38 Figure 3.8.2-10: ECCS Trip/Reset Actuator Valve Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-39 Figure 3.8.4-1: Reactor Building Concrete Structural Sections at First Floor (EL. 24'-0") . . . . . . 3.8-101 Figure 3.8.4-2: Reactor Building Concrete Structural Sections at Second Floor (EL. 50'-0") . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-102 Figure 3.8.4-3: Reactor Building Concrete Structural Sections at Third Floor (EL. 75'-0") . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-103 Figure 3.8.4-4: Reactor Building Concrete Structural Sections at Fourth Floor (EL. 100'-0"). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-104 Figure 3.8.4-5: Reactor Building Concrete Structural Sections at Fifth Floor (EL. 126'-0"). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-105 Figure 3.8.4-6: Reactor Building Concrete Structural Sections at RBC (EL. 145'-6") . . . . . . . . . . . 3.8-106 Figure 3.8.4-7: Reactor Building Concrete Structural Sections at Roof (EL. 181'-0") . . . . . . . . . . 3.8-107 Figure 3.8.4-8: Control Building Concrete Structural Sections at First Floor (EL. 50'-0"). . . . . . . 3.8-108 Figure 3.8.4-9: Control Building Concrete Structural Sections at Second Floor (EL. 76'-6") . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-109 Figure 3.8.4-10: Control Building Concrete Structural Sections at Third Floor (EL. 100'-0"). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-110 Tier 2 xxxi Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.8.4-11: Control Building Concrete Structural Sections at Fourth Floor (EL. 120'-0"). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-111 Figure 3.8.4-12: Control Building Steel Framing of Roof to EL. 141' 2" . . . . . . . . . . . . . . . . . . . . . . . . 3.8-112 Figure 3.8.4-13: East-West (X) Longitudinal Hydrodynamic Load Regions . . . . . . . . . . . . . . . . . . . . 3.8-113 Figure 3.8.4-14: North-South (Y) Transverse Hydrodynamic Load Regions . . . . . . . . . . . . . . . . . . . 3.8-114 Figure 3.8.4-15: Reactor Building SAP2000 Model (Looking Southwest). . . . . . . . . . . . . . . . . . . . . . 3.8-115 Figure 3.8.4-16: Elevation View of Reactor Building SAP2000 Model Looking South . . . . . . . . . . 3.8-116 Figure 3.8.4-17: Elevation View of Reactor Building SAP2000 Model Looking East . . . . . . . . . . . . 3.8-117 Figure 3.8.4-18: Longitudinal Section View of Reactor Building SAP2000 Model. . . . . . . . . . . . . . 3.8-118 Figure 3.8.4-19: Transverse Section View of Reactor Building SAP2000 Model. . . . . . . . . . . . . . . . 3.8-119 Figure 3.8.4-20: Reactor Building Exterior Walls with 7000 psi and 5000 psi Concrete . . . . . . . . 3.8-120 Figure 3.8.4-21: Control Building SAP2000 Model With Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-121 Figure 3.8.4-22: Control Building SAP2000 Model Without Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-122 Figure 3.8.4-23: Control Building SAP2000 Model View Looking West . . . . . . . . . . . . . . . . . . . . . . . 3.8-123 Figure 3.8.4-24: Control Building SAP2000 Model View Looking East . . . . . . . . . . . . . . . . . . . . . . . . 3.8-124 Figure 3.8.4-25: Control Building SAP2000 Model View Looking North . . . . . . . . . . . . . . . . . . . . . . 3.8-125 Figure 3.8.4-26: Control Building SAP2000 Model View Looking South . . . . . . . . . . . . . . . . . . . . . . 3.8-126 Figure 3.8.4-27: Total Static Lateral Soil Pressure Distribution Reactor Building . . . . . . . . . . . . . . 3.8-127 Figure 3.8.4-28: Seismic Soil Pressures on CRB Walls of Standalone Model . . . . . . . . . . . . . . . . . . . 3.8-128 Figure 3.8.4-29: Seismic Soil Pressures on CRB Walls of Triple Building Model . . . . . . . . . . . . . . . . 3.8-129 Figure 3.8.4-30: Enveloping Seismic Soil Pressures on CRB Walls of Standalone and Triple Building Models . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-130 Figure 3.8.4-31: Seismic Soil Pressure on RXB Walls from Standalone Model . . . . . . . . . . . . . . . . . 3.8-131 Figure 3.8.4-32: Seismic Soil Pressure on RXB Walls from Triple Building Model . . . . . . . . . . . . . . 3.8-132 Figure 3.8.4-33: Enveloping Soil Pressure on RXB Walls by Standalone and Triple Building Models . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-133 Figure 3.8.4-34: Reactor Flange Tool Support and Stand . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-134 Figure 3.8.4-35: Reactor Flange Tool Upper Support Wall Embed Plate . . . . . . . . . . . . . . . . . . . . . . 3.8-135 Figure 3.8.4-36: Reactor Flange Tool Base Embed Plate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-136 Figure 3.8.5-1: SAP2000 Model for Evaluation of Design Forces in the Reactor Building Basemat Model (X Axis is in the Longitudinal Direction, Y Axis is in the Transverse Direction, and Z Axis in the Vertical Upward Direction) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-184 Tier 2 xxxii Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.8.5-2: Static Base Pressure Contours for American Concrete Institute Load Combination 9-6 in the Reactor Building Basemat (psi) (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-185 Figure 3.8.5-2a: Static Base Pressure Contours for American Concrete Institute Load Combination 9-6 in the Control Building Basemat (psi) (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-186 Figure 3.8.5-3: Seismic Base Pressure Contours from SASSI2010 Analysis in the Reactor Building Basemat (psi) (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . 3.8-187 Figure 3.8.5-3a: Dynamic Pressure Contours on Control Building Basemat (psi) (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-188 Figure 3.8.5-4: Myy Due to Static Base Pressure on Reactor Building Basemat (kip-ft/ft) in the Reactor Building Basemat Model (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-189 Figure 3.8.5-4a: Myy Due to Static Loads on Control Building Basemat, Stand-Alone SAP2000 Model (kip-ft/ft) (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-190 Figure 3.8.5-5: Mxx Due to Static Base Pressure on Reactor Building Basemat (kip-ft/ft) in the Reactor Building Basemat Model (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-191 Figure 3.8.5-5a: Mxx Due to Static Loads on Control Building Basemat, Stand-Alone SAP2000 Model (kip-ft/ft) (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-192 Figure 3.8.5-6: Myy Due to Seismic Base Pressure on Reactor Building Basemat (kip-ft/ft) in the Reactor Building Basemat Model (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-193 Figure 3.8.5-6a: Myy Due to Seismic Base Pressure on Control Building Basemat (kip-ft/ft) (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-194 Figure 3.8.5-7: Mxx Due to Seismic Base Pressure on Reactor Building Basemat (kip-ft/ft) in the Reactor Building Basemat Model (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-195 Figure 3.8.5-7a: Mxx Due to Seismic Base Pressure on Control Building Basemat (kip-ft/ft) (Positive X Axis is to the Right of the Image and Positive Y is to the Top of the Image) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-196 Tier 2 xxxiii Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.8.5-8: RXB ANSYS Model with Backfill Soil. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-197 Figure 3.8.5-9: Nonlinear Contact Region between Building and Soil . . . . . . . . . . . . . . . . . . . . . . . 3.8-198 Figure 3.8.5-10: Edge and Center Nodes at Bottom of Foundations Selected for Building Settlement Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-199 Figure 3.8.5-11: RXB Skin Nodes on Backfill Soil Vertical Boundaries for Applying SASSI Acceleration Time Histories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-200 Figure 3.8.5-12: RXB Foundation Bottom Skin Nodes for Applying SASSI Acceleration Time Histories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-201 Figure 3.8.5-13: Displacements from SASSI Applied to ANSYS Model Boundary . . . . . . . . . . . . . . 3.8-202 Figure 3.8.5-14: Displacements from SASSI Applied to ANSYS Model Boundary . . . . . . . . . . . . . . 3.8-203 Figure 3.8.5-15: Nonlinear Contact Element between Backfill and Surrounding Soil . . . . . . . . . . 3.8-204 Figure 3.8.5-16: Buoyancy Load on Basemat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-205 Figure 3.8.5-17: Soil Type 7 - Acceleration Time History - Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-206 Figure 3.8.5-18: Soil Type 7 - Acceleration Time History - E-W . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-207 Figure 3.8.5-19: Soil Type 7 - Acceleration Time History - N-S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-208 Figure 3.8.5-20: Soil Type 8 - Acceleration Time History - Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-209 Figure 3.8.5-21: Soil Type 8 - Acceleration Time History - E-W . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-210 Figure 3.8.5-22: Soil Type 8 - Acceleration Time History - N-S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-211 Figure 3.8.5-23: Soil Type 11 - Acceleration Time History - Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-212 Figure 3.8.5-24: Soil Type 11 - Acceleration Time History - E-W . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-213 Figure 3.8.5-25: Soil Type 11 - Acceleration Time History - N-S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-214 Figure 3.8.5-26: Nonlinear Contact Region between CRB and Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-215 Figure 3.8.5-27: CRB Time Histories from SASSI Applied to ANSYS Model Boundary . . . . . . . . . . 3.8-216 Figure 3.8.5-28: Soil Type 11, Capitola Input - Acceleration Time History - Vertical. . . . . . . . . . . . 3.8-217 Figure 3.8.5-29: Soil Type 11, Capitola Input - Acceleration Time History - E-W . . . . . . . . . . . . . . . 3.8-218 Figure 3.8.5-30: Soil Type 11, Capitola Input - Acceleration Time History - N-S. . . . . . . . . . . . . . . . 3.8-219 Figure 3.8.5-31: Soil Type 7, Capitola Input - Acceleration Time History - Vertical . . . . . . . . . . . . . 3.8-220 Figure 3.8.5-32: Soil Type 7, Capitola Input - Acceleration Time History - E-W . . . . . . . . . . . . . . . . 3.8-221 Figure 3.8.5-33: Soil Type 7, Capitola Input - Acceleration Time History - N-S . . . . . . . . . . . . . . . . . 3.8-222 Figure 3.8.5-34: CRB Skin Nodes on Backfill Outer Boundaries for Applying SASSI Time Histories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-223 Figure 3.8.5-35: CRB Foundation Bottom Skin Nodes for Applying SASSI Time Histories . . . . . . 3.8-224 Figure 3.8.5-36: Buoyancy Load on Basemat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-225 Tier 2 xxxiv Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.8.5-37: Static Soil Pressure on CRB Outer Walls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-226 Figure 3.8.5-38: CRB Static Soil Pressure from Poisson's Ratio Effect - Soil Type 11 . . . . . . . . . . . . 3.8-227 Figure 3.8.5-39: CRB Static Soil Pressure from Poisson's Ratio Effect - Soil Type 7 . . . . . . . . . . . . . 3.8-228 Figure 3.8.5-40: CRB SAP2000 Model with Backfill Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-229 Figure 3.8.5-41: SAP2000 Model for Settlement. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-230 Figure 3.8.5-42: Total Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-231 Figure 3.8.5-43: Total Uncracked Base Vertical Reaction Time History due to Capitola for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-232 Figure 3.8.5-44: Total Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-233 Figure 3.8.5-45: Total Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-234 Figure 3.8.5-46: Total Cracked Base Vertical Reaction Time History due to Lucerne for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-235 Figure 3.8.5-47: Total Cracked Base Vertical Reaction Time History due to Lucerne for Soil Type 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-236 Figure 3.8.5-48: CRB Foundation Time History Location Designations . . . . . . . . . . . . . . . . . . . . . . . 3.8-237 Figure 3.8.5-49: Reaction Force at Location A (S11 - Vertical Excitation) . . . . . . . . . . . . . . . . . . . . . . 3.8-238 Figure 3.8.5-50: Relative Displacement (Uplift) at Location A (S11 - Vertical Excitation) . . . . . . . 3.8-239 Figure 3.8.5-51: Lateral Relative Displacements (Sliding) at Location A (S11 - Vertical Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-240 Figure 3.8.5-52: RXB Foundation Time History Location Designations . . . . . . . . . . . . . . . . . . . . . . . 3.8-241 Figure 3.8.5-53: Lateral Relative Displacements (Sliding) at Location A (S7 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-242 Figure 3.8.5-54: Lateral Relative Displacements (Sliding) at Location B (S7 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-243 Figure 3.8.5-55: Lateral Relative Displacements (Sliding) at Location C (S7 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-244 Figure 3.8.5-56: Lateral Relative Displacements (Sliding) at Location D (S7 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-245 Figure 3.8.5-57: Lateral Relative Displacements (Sliding) at Location A (S7 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-246 Figure 3.8.5-58: Lateral Relative Displacements (Sliding) at Location B (S7 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-247 Figure 3.8.5-59: Lateral Relative Displacements (Sliding) at Location C (S7 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-248 Tier 2 xxxv Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.8.5-60: Lateral Relative Displacements (Sliding) at Location D (S7 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-249 Figure 3.8.5-61: Lateral Relative Displacements (Sliding) at Location A (S11 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-250 Figure 3.8.5-62: Lateral Relative Displacements (Sliding) at Location B (S11 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-251 Figure 3.8.5-63: Lateral Relative Displacements (Sliding) at Location C (S11 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-252 Figure 3.8.5-64: Lateral Relative Displacements (Sliding) at Location D (S11 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-253 Figure 3.8.5-65: Lateral Relative Displacements (Sliding) at Location A (S11 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-254 Figure 3.8.5-66: Lateral Relative Displacements (Sliding) at Location B (S11 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-255 Figure 3.8.5-67: Lateral Relative Displacements (Sliding) at Location C (S11 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-256 Figure 3.8.5-68: Lateral Relative Displacements (Sliding) at Location D (S11 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-257 Figure 3.8.5-69: Lateral Relative Displacements (Sliding) at Location A (S8 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-258 Figure 3.8.5-70: Lateral Relative Displacements (Sliding) at Location B (S8 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-259 Figure 3.8.5-71: Lateral Relative Displacements (Sliding) at Location C (S8 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-260 Figure 3.8.5-72: Lateral Relative Displacements (Sliding) at Location D (S8 - E-W Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-261 Figure 3.8.5-73: Lateral Relative Displacements (Sliding) at Location A (S8 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-262 Figure 3.8.5-74: Lateral Relative Displacements (Sliding) at Location B (S8 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-263 Figure 3.8.5-75: Lateral Relative Displacements (Sliding) at Location C (S8 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-264 Figure 3.8.5-76: Lateral Relative Displacements (Sliding) at Location D (S8 - N-S Excitation). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-265 Figure 3.8.5-77: Total CRB Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-266 Figure 3.8.5-78: Total CRB Uncracked Base Vertical Reaction Time History due to Capitola for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-267 Tier 2 xxxvi Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3.8.5-79: Total CRB Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-268 Figure 3.8.5-80: Total CRB Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-269 Figure 3.8.5-81: Total CRB Cracked Base Vertical Reaction Time History due to Lucerne for Soil Type 7. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-270 Figure 3.8.5-82: Total CRB Cracked Base Vertical Reaction Time History due to Lucerne for Soil Type 9. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-271 Figure 3.9-1: Nuscale Power Module Showing Reactor Vessel Internals Component Assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-103 Figure 3.9-2: Upper Riser Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-104 Figure 3.9-3: Lower Riser Assembly. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-105 Figure 3.9-4: Core Support Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-106 Figure 3B-1: Whitney Rectangular Stress Block . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-189 Figure 3B-2: SAP2000 Membrane and Shear Force Definition . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-190 Figure 3B-3: SAP2000 Bending Moment Definition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-191 Figure 3B-4: SASSI2010 Membrane, Shear Force, and Bending Moment Definitions . . . . . . 3B-192 Figure 3B-5: SAP2000 Frame Element Results Definition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-193 Figure 3B-6: SASSI2010 Frame Element Results Definition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-194 Figure 3B-7: SAP2000 Elevation View and Shell Element Numbers at RXB Grid Line 1 (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-195 Figure 3B-8: RXB Reinforcement Elevation at Grid Line 1 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-196 Figure 3B-9: RXB Reinforcement Section View of Wall on Grid Line 1 . . . . . . . . . . . . . . . . . . . . . 3B-197 Figure 3B-10: SAP2000 Elevation View and Shell Element Numbers at RXB Grid Line 3 (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-198 Figure 3B-11: RXB Reinforcement Elevation at Grid Line 3 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-199 Figure 3B-12: RXB Reinforcement Section View of Pool Weir Wall on Grid Line 3 . . . . . . . . . . . 3B-200 Figure 3B-13: RXB Reinforcement Section View of Stiffener Wall on Grid Line 3 . . . . . . . . . . . . 3B-201 Figure 3B-14: SAP2000 Elevation View and Shell Element Numbers at RXB Grid Line 4 (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-202 Figure 3B-15: RXB Reinforcement Elevation at Grid Line 4 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-203 Figure 3B-16: RXB Reinforcement Section View of 5 ft Thick Wall on Grid Line 4. . . . . . . . . . . . 3B-204 Figure 3B-17: RXB Reinforcement Section View of 4 ft Thick Wall on Grid Line 4. . . . . . . . . . . . 3B-205 Figure 3B-18: SAP2000 Elevation View and Shell Element Numbers at RXB Grid Line 6 (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-206 Tier 2 xxxvii Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3B-19: RXB Reinforcement Elevation at Grid Line 6 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-207 Figure 3B-20: RXB Reinforcement Section View of Upper Stiffener Wall on Grid Line 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-208 Figure 3B-21: RXB Reinforcement Section Views of Pool Wall on Grid Line 6 . . . . . . . . . . . . . . . 3B-209 Figure 3B-22: SAP2000 Elevation View and Shell Element Numbers at RXB Grid Line E (Looking North) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-210 Figure 3B-23: RXB Reinforcement Elevation at Grid Line E Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-211 Figure 3B-24: RXB Reinforcement Section View of Wall on Grid Line E . . . . . . . . . . . . . . . . . . . . . 3B-212 Figure 3B-25: SAP2000 Plan View and Shell Element Numbers on Slab at RXB EL 100-0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-213 Figure 3B-26: RXB Reinforcement Plan at EL 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-214 Figure 3B-27: RXB Reinforcement Section View of Slab at EL 100'-0" . . . . . . . . . . . . . . . . . . . . . . . 3B-215 Figure 3B-28: SAP2000 Plan View and Shell Element Numbers on RXB Roof Slab. . . . . . . . . . . 3B-216 Figure 3B-29: RXB Reinforcement Plan for Roof Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-217 Figure 3B-30: RXB Reinforcement Section View of Roof Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-218 Figure 3B-31: SAP2000 View and Frame Element Numbers of Pilasters on RXB Grid Line A Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-219 Figure 3B-32: RXB Reinforcement Detail for Pilaster Type 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-220 Figure 3B-33: RXB Reinforcement Detail for Pilaster Type 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-221 Figure 3B-34: RXB Reinforcement Detail for Pilaster Type 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-222 Figure 3B-35: RXB Reinforcement Detail for Pilaster Type 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-223 Figure 3B-36: RXB Reinforcement Detail for Pilaster Type 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-224 Figure 3B-37: SAP2000 View and Frame Element Numbers of Beams on RXB EL 75'-0" Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-225 Figure 3B-38: RXB Reinforcement Detail for Type 1 T-Beams at EL 75'-0" . . . . . . . . . . . . . . . . . . . 3B-226 Figure 3B-39: RXB Reinforcement Detail for Type 2 T-Beams at EL 75'-0" . . . . . . . . . . . . . . . . . . . 3B-227 Figure 3B-40: SAP2000 View and Frame Element Numbers of Buttresses at Grid Line 1 on RXB EL. 126'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-228 Figure 3B-41: RXB Reinforcement Detail for Buttress Type 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-229 Figure 3B-42: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-230 Figure 3B-43: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-231 Figure 3B-44: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-232 Figure 3B-45: SAP2000 Elevation View and Shell Element Numbers at RXB Wall at Grid Line B (Looking North) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-233 Tier 2 xxxviii Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3B-46: RXB Reinforcement Elevation at RXB Wall at Grid Line B . . . . . . . . . . . . . . . . . . . . . 3B-234 Figure 3B-47: RXB Reinforcement Section View of RXB Wall at Grid Line B . . . . . . . . . . . . . . . . . 3B-235 Figure 3B-48: NuScale Power Module Base Support Assembly at Reactor Building Pool Floor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-236 Figure 3B-49: Plan View and Cross Sections of NPM Embed Plate. . . . . . . . . . . . . . . . . . . . . . . . . . 3B-237 Figure 3B-50: Plan View of NPM Embed Plate Anchorage and Passive Plate. . . . . . . . . . . . . . . . 3B-238 Figure 3B-51: NPM Lug Support Plan View and Details. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-239 Figure 3B-52: NPM Lug Location . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-240 Figure 3B-53: NPM Lug Support SAP2000 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-241 Figure 3B-54: NPM Lug Support SAP2000 Model Close-Up . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-242 Figure 3B-55: NPM Lug Support Liner Plate Section. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-243 Figure 3B-56: NPM Lug Support Liner Plate and Shear Lugs (Shown in Red) . . . . . . . . . . . . . . . 3B-244 Figure 3B-57: NPM Lug Support Model showing internal Stiffener Plates . . . . . . . . . . . . . . . . . . 3B-245 Figure 3B-58: NPM Lug Support Loading (W-Lug-PY+). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-246 Figure 3B-59: NPM Lug Support Loading (W-Lug-PY-) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-247 Figure 3B-60: NPM Lug Support SAP2000 Model Restraints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-248 Figure 3B-61: Stiffener Plate Section Cut Groups (Fins). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-249 Figure 3B-62: S11 Stress plotted on the Deflected Shape due to Load Combination W-Lug-PY+ (psi) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-250 Figure 3B-63: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-251 Figure 3B-64: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-252 Figure 3B-65: SAP2000 Elevation View and Shell Element Numbers at CRB Grid Line 3 (Looking North) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-253 Figure 3B-66: CRB Reinforcement Elevation at Grid Line 3 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-254 Figure 3B-67: CRB Reinforcement Section View of Wall on Grid Line 3 . . . . . . . . . . . . . . . . . . . . . 3B-255 Figure 3B-68: SAP2000 Elevation View and Shell Element Numbers at CRB Grid Line 4 (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-256 Figure 3B-69: CRB Reinforcement Elevation at Grid Line 4 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-257 Figure 3B-70: CRB Reinforcement Section View of Wall on Grid Line 4 . . . . . . . . . . . . . . . . . . . . . 3B-258 Figure 3B-71: SAP2000 Elevation View and Shell Element Numbers at Grid Line A (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-259 Figure 3B-72: CRB Reinforcement Elevation at Grid Line A Wall. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-260 Figure 3B-73: CRB Reinforcement Section View of Wall on Grid Line A . . . . . . . . . . . . . . . . . . . . . 3B-261 Figure 3B-74: CRB Basemat View of Finite Element Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-262 Tier 2 xxxix Revision 4

NuScale Final Safety Analysis Report List of Figures LIST OF FIGURES Figure 3B-75: CRB Reinforcement Plan of Basemat Foundation . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-263 Figure 3B-76: Cross Section of CRB Basemat Showing Reinforcing Steel . . . . . . . . . . . . . . . . . . . 3B-264 Figure 3B-77: SAP2000 Plan View and Shell Element Numbers on CRB Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-265 Figure 3B-78: CRB Reinforcement Plan at EL. 100-0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-266 Figure 3B-79: CRB Reinforcement Section Views of Slab at EL. 100-0 . . . . . . . . . . . . . . . . . . . . . 3B-267 Figure 3B-80: SAP2000 View and Frame Element Numbers of Pilasters on CRB Grid Line 1 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-268 Figure 3B-81: CRB Reinforcement Detail for Pilaster Type 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-269 Figure 3B-82: CRB Reinforcement Detail for Pilaster Type 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-270 Figure 3B-83: SAP2000 View and Frame Element Numbers of T-Beams on CRB EL. 120'-0" Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-271 Figure 3B-84: CRB Reinforcement Detail for T-Beam (Type 1) at EL. 120'-0". . . . . . . . . . . . . . . . . 3B-272 Figure 3B-85: CRB Reinforcement Detail for T-Beam (Type 2) at EL. 120'-0". . . . . . . . . . . . . . . . . 3B-273 Figure 3B-86: Reactor Building Basemat Perimeter Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-274 Figure 3B-87: Reactor Building Basemat Interior Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-275 Figure 3B-88: Reactor Building Reinforcement Plan of Basemat Foundation . . . . . . . . . . . . . . . 3B-276 Figure 3B-89: Cross Section of Reactor Building Basemat Showing Reinforcing Steel. . . . . . . 3B-277 Figure 3C-1: Containment Liquid Space Metal and Liquid Temperatures with Bounding Curve (Zones A and B) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-30 Figure 3C-2: Containment Vapor Space Metal and Gas Temperatures with Bounding Curve (Zones C, D, E, and F) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-31 Figure 3C-3: Bounding Envelope for Average Vapor Temperature at Top of Module (Zone G). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-32 Figure 3C-4: Bounding Envelope for Maximum Vapor Temperatures at Reactor Building El 145'-0 (Zone H) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-33 Tier 2 xl Revision 4

Conformance with U.S. Nuclear Regulatory Commission General Design Criteria This section addresses design compliance with the General Design Criteria (GDC) in 10 CFR 50, Appendix A, for safety-related and when appropriate, risk-significant structures, systems, and components (SSC).

The following sections state the criterion and then address how the criterion is implemented in the NuScale Power Plant design. The section provides a statement regarding the conformance or exception, as well as a list of sections where additional information on conformance is presented.

In certain cases, NuScale meets the intent of the GDC or has developed a principal design criterion (PDC) to address the specific design of the NuScale Power Plant pressurized water reactor.

1 Overall Requirements 1.1 Criterion 1-Quality Standards and Records Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function. A quality assurance program shall be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety functions.

Appropriate records of the design, fabrication, erection, and testing of structures, systems, and components important to safety shall be maintained by or under the control of the nuclear power unit licensee throughout the life of the unit.

Implementation in the NuScale Power Plant Design NuScale's quality assurance (QA) program satisfies the requirements of 10 CFR 50 Appendix B and ASME NQA-1-2008 and NQA-1a-2009 addenda, "Quality Assurance Requirements for Nuclear Facility Applications" (Reference 3.1-1). As such, the NuScale QA program provides confidence that the SSC that are required to perform safety-related and risk-significant functions will perform the functions satisfactorily.

NuScale's QA program is described in the NuScale Quality Assurance Program Description (QAPD).

NuScale plant SSC are assigned safety and QA classifications based on their safety and risk-significant functions. The QA classification is used to identify and apply appropriate QA requirements for safety-related and risk-significant SSC. The safety and QA classifications assigned to NuScale plant SSC are indicated in Table 3.2-1.

2 3.1-1 Revision 4

erection, and testing and maintained throughout the life of the plant.

Conformance or Exception The NuScale Power Plant design conforms to GDC 1.

Relevant FSAR Chapters and Sections Section 3.2 Classification of Structures, Systems, and Components Section 3.9 Mechanical Systems and Components Section 3.10 Seismic and Dynamic Qualifications of Mechanical and Electrical Equipment Section 3.11 Environmental Qualification of Mechanical and Electrical Equipment Section 3.13 Threaded Fasteners (ASME Code Class 1, 2, and 3)

Chapter 5 Reactor Coolant System and Connecting Systems Chapter 6 Engineered Safety Features Chapter 7 Instrumentation and Controls Section 9.1.5 Overhead Heavy Load Handling System Section 9.3 Process Auxiliaries Chapter 17 Quality Assurance and Reliability Assurance 1.2 Criterion 2-Design Bases for Protection Against Natural Phenomena Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems, and components shall reflect: (1) Appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period of time in which the historical data have been accumulated, (2) Appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena and (3) the importance of the safety functions to be performed.

Implementation in the NuScale Power Plant Design The safety-related SSC in the NuScale Power Plant are designed to withstand the effects of natural phenomena based on parameters selected to bound the hazardous 2 3.1-2 Revision 4

including appropriate combinations of the effects of normal operating and accident conditions. The NuScale Power Plant's site parameters are listed in Table 2.0-1. Seismic and quality group classifications, and other pertinent standards and information are provided in Table 3.2-1.

Conformance or Exception The NuScale Power Plant design conforms to GDC 2.

Relevant FSAR Chapters and Sections Chapter 2 Site Characteristics and Site Parameters Section 3.2 Classification of Structures, Systems, and Components Section 3.3 Wind and Tornado Loadings Section 3.4 Water Level (Flood) Design Section 3.5 Missile Protection Section 3.7 Seismic Design Section 3.8 Design of Category I Structures Section 3.9 Mechanical Systems and Components Section 3.10 Seismic and Dynamic Qualifications of Mechanical and Electrical Equipment Section 3.11 Environmental Qualification of Mechanical and Electrical Equipment Section 3.12 ASME Code Class 1, 2, and 3 Piping Systems, Piping Components and Associated Supports Chapter 5 Reactor Coolant System and Connecting Systems Chapter 6 Engineered Safety Features Section 7.1 Fundamental Design Principles Section 8.3 Onsite Power Systems Section 9.1.2 New and Spent Fuel Storage Section 9.1.3 Spent Fuel Pool Cooling and Cleanup System Section 9.3 Process Auxiliaries 2 3.1-3 Revision 4

Chapter 15 Transient and Accident Analyses 1.3 Criterion 3-Fire Protection Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

Implementation in the NuScale Power Plant Design The NuScale Power Plant fire protection design and program ensure that the requirements of 10 CFR 50.48 and GDC 3 are met. The SSC are designed and located to minimize the probability and effects of fires and explosions. Noncombustible and fire-resistant materials are used throughout the plant where fire is a potential risk to safety-related systems. Fire barriers ensure that redundant, safety-related systems and components are separated to assure that a fire in one area will not affect the redundant systems and components in an adjacent area from performing their safety functions.

Buildings that contain equipment required for safe shutdown are compartmentalized to minimize the impacts of a fire. These divisions and sub-divisions ensure adequate equipment and cable separation meet the enhanced fire protection criteria.

Compartmentalization is achieved by using properly rated fire barriers, fire doors, fire dampers, and penetration seals to prevent the spread of fire between areas.

The fire protection system and equipment is designed in accordance with the guidance provided in Regulatory Guide 1.189, Revision 2, and applicable National Fire Protection Association codes. This ensures that the fire detection and fighting systems provided have the capacity and capability to minimize the adverse effects of fires and that their rupture or inadvertent operation does not significantly impair the safety capability of other SSC.

Conformance or Exception The NuScale Power Plant design conforms to GDC 3.

Relevant FSAR Chapters and Sections Section 9.3 Process Auxiliaries Section 9.4 Air Conditioning, Heating, Cooling, and Ventilation Systems 2 3.1-4 Revision 4

Appendix 9A Fire Hazard Analysis Section 11.2 Liquid Waste Management System Section 11.3 Gaseous Radioactive Waste Management System 1.4 Criterion 4-Environmental and Dynamic Effects Design Bases Structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit. However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

Implementation in the NuScale Power Plant Design The design of safety-related and risk-significant SSC is such that the effects of environmental conditions associated with normal operation, maintenance testing, and postulated accidents, including LOCAs, are accommodated. The NuScale Power Plant design appropriately protects against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the NuScale Power Module (NPM) and prevents piping failure using leak-before-break methodology.

Conformance or Exception The NuScale Power Plant design conforms to GDC 4.

Relevant FSAR Chapters and Sections Section 3.3 Wind and Tornado Loadings Section 3.4 Water Level (Flood) Design Section 3.5 Missile Protection Section 3.6 Protection against Dynamic Effects Associated with Postulated Rupture of Piping Section 3.8 Design of Category I Structures Section 3.9 Mechanical Systems and Components 2 3.1-5 Revision 4

Section 3.11 Environmental Qualification of Mechanical and Electrical Equipment Section 3.12 ASME Code Class 1, 2, and 3 Piping Systems, Piping Components and Associated Supports Section 3.13 Threaded Fasteners (ASME Code Class 1, 2, and 3)

Section 4.6 Functional Design of Control Rod Drive System Chapter 5 Reactor Coolant System and Connecting Systems Chapter 6 Engineered Safety Features Chapter 7 Instrumentation and Controls Section 8.3 Onsite Power Systems Chapter 9 Auxiliary Systems Chapter 10 Steam and Power Conversion System Chapter 15 Transient and Accident Analyses 1.5 Criterion 5-Sharing of Structures, Systems, and Components Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units.

Implementation in the NuScale Power Plant Design The term NuScale Power Plant refers to the entire site, including up to 12 NPMs and the associated balance of plant support systems and structures. The design considers the safety effects and the risk associated with multi-module plant operation with shared or common systems such that each NPM can be safely operated independent of other NPMs. The plant includes design features that ensure the independence and protection of NPM safety systems during all operational modes. Given a single failure in safety-related SSC in one NPM, these design features ensure that safety functions are capable of being performed in other NPMs. The NuScale Power Plant is designed such that a failure of a shared system, which are nonsafety-related with exception of the ultimate heat sink (UHS), does not prevent the performance of NPM safety functions.

Conformance or Exception The NuScale Power Plant design conforms to GDC 5.

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Section 5.4.3 Decay Heat Removal System Section 6.2 Containment Systems Section 6.3 Emergency Core Cooling System Section 6.4 Control Room Habitability Chapter 7 Instrumentation and Controls Chapter 8 Electric Power Chapter 9 Auxiliary Systems Chapter 10 Steam and Power Conversion System Chapter 15 Transient and Accident Analyses Chapter 21 Multi-Module Design Considerations 2 Protection by Multiple Fission Product Barriers 2.1 Criterion 10-Reactor Design The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences.

Implementation in the NuScale Power Plant Design The reactor core and associated coolant, control, and protection systems are designed with appropriate margin such that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences (AOOs).

During AOOs and low probability events that may result in a plant shutdown, the NuScale Power Plant is designed such that the reactor will be brought to subcritical conditions and maintained in safe shutdown. The reactor core is designed to maintain integrity over a complete range of power levels and sized with sufficient heat transfer area and coolant flow such that SAFDLs are not exceeded.

Safety analysis design limits are established to demonstrate conformance with GDC 10.

These limits ensure that the fuel boundary is not breached, thus leaving the first fission product barrier intact. SAFDLs also ensure that the fuel system dimensions remain within operational tolerances and that the functional capabilities are not reduced below those assumed in the safety analysis.

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The NuScale Power Plant design conforms to GDC 10.

Relevant FSAR Chapters and Sections Section 3.9.5 Reactor Vessel Internals Section 4.2 Fuel System Design Section 4.3 Nuclear Design Section 4.4 Thermal and Hydraulic Design Chapter 7 Instrumentation and Controls Section 9.3.4 Chemical and Volume Control System Chapter 15 Transient and Accident Analyses 2.2 Criterion 11-Reactor Inherent Protection The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tends to compensate for a rapid increase in reactivity.

Implementation in the NuScale Power Plant Design The reactor core and associated coolant systems are designed such that inherent reactivity control is provided during changing plant conditions. The two main feedback effects that compensate for a rapid increase in reactivity are the fuel Doppler temperature reactivity coefficient and the fuel moderator temperature coefficient.

Conformance or Exception The NuScale Power Plant design conforms to GDC 11.

Relevant FSAR Chapters and Sections Section 4.3 Nuclear Design 2.3 Criterion 12-Suppression of Reactor Power Oscillations The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily detected and suppressed.

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The NuScale reactor core is designed to assure that power oscillations, which can result in conditions exceeding SAFDLs, are not possible. Oscillations are evaluated at the beginning, middle, and end of the equilibrium cycle. The NuScale reactor core is stable with respect to axial and radial stability, as discussed in Section 4.3.2.

Oscillations in core power can be readily detected by the fixed in-core detector system, which continuously monitors the core flux distribution.

The reactor core and associated coolant, control, and protection systems ensure that power and hydraulic oscillations that can result in conditions exceeding SAFDLs are not possible. Hydraulic stability protection is achieved by the regional exclusion method.

The module protection system (MPS) enforces this regional exclusion by ensuring the NPM maintains adequate riser subcooling.

Conformance or Exception The NuScale Power Plant design conforms to GDC 12.

Relevant FSAR Chapters and Sections Section 4.3 Nuclear Design Section 4.4 Thermal and Hydraulic Design Section 15.9 Stability 2.4 Criterion 13-Instrumentation and Control Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

Implementation in the NuScale Power Plant Design Instrumentation and controls are provided to monitor variables and systems over their anticipated ranges for normal operations, AOOs, and postulated accident conditions to assure adequate safety. The design of the NuScale safety-related instrument and control systems is based on independence, redundancy, predictability and repeatability, and diversity and defense-in-depth. The appropriate controls are provided to the NPM with sufficient margin to ensure these variables and systems remain within the prescribed operating ranges.

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The NuScale Power Plant design conforms to GDC 13.

Relevant FSAR Chapters and Sections Chapter 6 Engineered Safety Features Chapter 7 Instrumentation and Controls Chapter 9 Auxiliary Systems Chapter 15 Transient and Accident Analyses 2.5 Criterion 14-Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Implementation in the NuScale Power Plant Design The reactor pressure vessel (RPV) and pressure retaining components associated with the reactor coolant pressure boundary (RCPB) are designed and fabricated with sufficient margin to assure the RCPB behaves in a non-brittle manner and to minimize the probability of abnormal leakage, rapidly propagating fracture, and gross rupture.

The RCPB materials meet the fabrication, construction, and testing requirements of the ASME Boiler and Pressure Vessel Code (BPVC),Section III Division 1, Subsection NB (Reference 3.1-2) and the materials selected for fabrication of the RCPB meet the ASME BPVC,Section II (Reference 3.1-3) requirements.

The primary and secondary water chemistry, along with the water chemistry for the pools forming the ultimate heat sink, is controlled to monitor for chemical species that can affect the RCPB integrity. Sampling and analysis of reactor coolant and pool water samples verify that key chemistry parameters are within prescribed limits and that impurities are properly controlled. This provides assurance that corrosion is mitigated and will not adversely affect the RCPB.

Conformance or Exception The NuScale Power Plant design conforms to GDC 14.

Relevant FSAR Chapters and Sections Section 3.9 Mechanical Systems and Components Section 3.12 ASME Code Class 1, 2, and 3 Piping Systems, Piping Components, and Associated Supports Section 3.13 Threaded Fasteners (ASME Code Class 1, 2, and 3) 2 3.1-10 Revision 4

Section 9.3 Process Auxiliaries Section 10.3.5 Water Chemistry Section 10.4.6 Condensate Polishing System 2.6 Criterion 15-Reactor Coolant System Design The reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.

Implementation in the NuScale Power Plant Design The overpressure protection system is designed with sufficient capacity to prevent the RCPB from exceeding 110 percent of design pressure during normal operations and AOOs. The system ensures that design limits are not exceeded during an anticipated transient without scram. The overpressure protection system is able to perform its function assuming a single active failure and concurrent loss of normal AC power.

Overpressure protection is provided by the reactor safety valves and in accordance with the requirements of ASME Code,Section III Division 1, Subsection NB for the RCPB and Subsection NC (Reference 3.1-4) for the secondary side of the steam generator and decay heat removal system (DHRS).

Conformance or Exception The NuScale Power Plant design conforms to GDC 15.

Relevant FSAR Chapters and Sections Section 3.9 Mechanical Systems and Components Section 3.12 ASME Code Class 1, 2, and 3 Piping Systems, Piping Components and Associated Supports Chapter 5 Reactor Coolant System and Connecting Systems Chapter 7 Instrumentation and Controls Chapter 15 Transient and Accident Analyses 2.7 Criterion 16-Containment Design Reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the 2 3.1-11 Revision 4

Implementation in the NuScale Power Plant Design The containment and associated systems are designed to establish an essentially leak-tight barrier against an uncontrolled release of radioactivity to the environment, and assures that containment design conditions are not exceeded for as long as the postulated accident conditions require. The integrity of the containment vessel (CNV) and the passive isolation barriers, along with the isolation of the lines that penetrate primary containment accomplish the provisions of GDC 16.

Conformance or Exception The NuScale Power Plant design conforms to GDC 16.

Relevant FSAR Chapters and Sections Section 3.8.2 Steel Containment Section 6.2 Containment Systems 2.8 Criterion 17-Electric Power Systems An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function for each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded as a result of anticipated operational occurrences and (2) the core is cooled and containment integrity and other vital functions are maintained in the event of postulated accidents.

The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure.

Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to both circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all onsite alternating current power supplies and the other offsite electric power circuit, to assure that specified acceptable fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, containment integrity, and other vital safety functions are maintained.

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generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies.

Implementation in the NuScale Power Plant Design The NuScale Power Plant is designed with passive safety-related systems for safe shutdown, core and spent fuel assembly cooling, containment isolation and integrity, and RCPB integrity. Electrical power is not relied upon to meet SAFDLs or to protect the RCPB as a result of AOOs or postulated accidents. The availability of electrical power sources does not affect the ability to achieve and maintain safety-related functions.

Although not relied on to ensure plant safety-related functions are achieved, the design of the AC and DC power systems includes provisions for independence and redundancy.

Conformance or Exception The NuScale Power Plant design does not conform to GDC 17. The NuScale design supports an exemption from the criterion.

Relevant FSAR Chapters and Sections Chapter 8 Electric Power Chapter 15 Transient and Accident Analyses 2.9 Criterion 18-Inspection and Testing of Electric Power Systems Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features, such as wiring, insulation, connections, and switchboards, to assess the continuity of the systems and the condition of their components. The systems shall be designed with a capability to test periodically (1) the operability and functional performance of the components of the systems, such as onsite power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into operation, including operation of applicable portions of the protection system, and the transfer of power among the nuclear power unit, the offsite power system, and the onsite power system.

Implementation in the NuScale Power Plant Design The electric power supply systems in the NuScale Power plant do not contain any safety-related or risk-significant SSC that are required to meet GDC 18. Although not relied on to meet GDC 18, the plant design does include provisions for testing and inspecting of power supply systems.

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The NuScale Power Plant design does not conform to GDC 18. The NuScale design supports an exemption from the criterion.

Relevant FSAR Chapters and Sections Chapter 8 Electric Power 2.10 Criterion 19-Control Room A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Applicants for and holders of construction permits and operating licenses under this part who apply on or after January 10, 1997, applicants for design approvals or certifications under part 52 of this chapter who apply on or after January 10, 1997, applicants for and holders of combined licenses or manufacturing licenses under part 52 of this chapter who do not reference a standard design approval or certification, or holders of operating licenses using an alternative source term under 50.67, shall meet the requirements of this except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv (5 rem) total effective dose equivalent (TEDE) as defined in 50.2 for the duration of the accident.

Implementation in the NuScale Power Plant Design The NuScale design supports an exemption from the provisions of GDC 19. The following PDC has been adopted:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) as defined in 10 CFR 50.2 for the duration of the accident.

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instrumentation and controls to maintain the unit in a safe shutdown condition.

The NuScale Power main control room contains the instrumentation and controls necessary to operate the NPMs safely under normal conditions and to maintain them in a safe condition under accident conditions, including a LOCA. Adequate protection is provided to permit access and occupancy of the control room so that personnel do not receive a whole body dose greater than 5 rem.

Heating, ventilation, and air conditioning are normally provided to the main control room by the control room ventilation system. Redundant toxic gas detectors, smoke detectors, and radiation detectors are provided in the outside air duct, upstream of both the control room ventilation system filter units and the bubble tight outdoor air isolation dampers. Upon detection of a high radiation level in the outside air intake, the system is realigned so that 100 percent of the outside air passes through the control room ventilation system filter unit. When power is unavailable, or if high levels of radiation are detected downstream of the charcoal filtration unit, the control room ventilation system filter unit is stopped, the outside air intake is automatically isolated, and the bubble-tight isolation dampers are closed. Once the control room envelope dampers are closed, the control room envelope is maintained for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by the control room habitability system.

The NuScale main control room (MCR) is designed with the ability to place the reactors in safe shutdown in the event of an MCR evacuation event, and for safe shutdown to be maintained without operator action thereafter. Prior to evacuating the MCR, operators trip the reactors, initiate decay heat removal and initiate containment isolation. These actions result in passive cooling that achieves safe shutdown of the reactors. Operators can also achieve safe shutdown of the reactors from outside the MCR in the MPS equipment rooms within the reactor building. Following shutdown and initiation of passive cooling from either the MCR or the MPS equipment rooms, the NuScale design does not rely on operator action, instrumentation, or controls outside of the MCR to maintain safe shutdown condition. The design includes a remote shutdown station (RSS) for monitoring of the plant if the MCR is evacuated. There are no displays, alarms, or controls in the RSS credited to meet the requirements of principal design criterion (PDC) 19 as there is no manual control of safety-related equipment allowed from the RSS.

Conformance or Exception The NuScale Power Plant design departs from GDC 19 and supports an exemption from the criterion. The NuScale Power Plant design conforms to PDC 19.

Relevant FSAR Chapters and Sections Section 5.4.3 Decay Heat Removal System Section 6.4 Control Room Habitability Section 7.1 Fundamental Design Principles 2 3.1-15 Revision 4

Section 9.5 Other Auxiliary Systems Appendix 9A Fire Hazard Analysis Section 11.5 Process and Effluent Radiation Monitoring Instrumentation and Sampling Section 12.3 Radiation Protection Design Features Chapter 15 Transient and Accident Analyses Section 18.7 Human-System Interface Design 3 Protection and Reactivity Control Systems 3.1 Criterion 20-Protection System Functions The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

Implementation in the NuScale Power Plant Design The MPS monitors process parameters that are directly related to equipment mechanical limitations, monitors parameters that directly affect the heat transfer capability of the NPM, and automatically executes safety-related functions in response to out-of-normal conditions. The MPS, in response to the NPM exceeding an analytical safety limit, trips the reactor. The MPS also actuates the engineered safety features actuation system (ESFAS) when specified setpoints are exceeded to prevent or mitigate damage to the reactor core and RCS.

Conformance or Exception The NuScale Power Plant design conforms to GDC 20.

Relevant FSAR Chapters and Sections Chapter 7 Instrumentation and Controls Chapter 15 Transient and Accident Analyses 3.2 Criterion 21-Protection System Reliability and Testability The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) 2 3.1-16 Revision 4

redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.

Implementation in the NuScale Power Plant Design The MPS incorporates the design principles of redundancy and independence such that no single failure results in the loss of the protective function. The MPS has four redundant groups of signal conditioning and trip determination, two divisions of reactor trip systems (RTSs) and ESFAS, and redundant communication paths. Each safety function uses two-out-of-four voting logic with two independent divisions of RTS and ESFAS so that a single failure will not prevent the safety function from being accomplished. The MPS SSC are designed to be tested and calibrated while retaining the capability to accomplish its required safety function. The MPS is designed for high functionality and to permit periodic testing during operation, including the ability to test channels independently to determine if failures or a loss of redundancy have occurred.

Conformance or Exception The NuScale Power Plant design conforms to GDC 21.

Relevant FSAR Chapters and Sections Chapter 7 Instrumentation and Controls Section 9.3.4 Chemical and Volume Control System 3.3 Criterion 22-Protection System Independence The protection system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function.

Implementation in the NuScale Power Plant Design The MPS equipment is located in the Reactor Building and is designed to enable systems and components required for safe plant operation to withstand natural phenomena, postulated design basis accidents, and design basis threats. The MPS has four redundant groups of signal conditioning and trip determination, two divisions of RTS and ESFAS, and redundant communication paths. Each safety function uses two-out-of-four voting logic with two independent divisions of RTS and ESFAS so that a 2 3.1-17 Revision 4

accomplish its required safety function. The MPS is designed for high functionality and to permit periodic testing during operation, including the ability to test channels independently to determine if failures or a loss of redundancy have occurred. To the extent practical, functional diversity and diversity in component design is used to perform the protection functions and prevent its loss.

Conformance or Exception The NuScale Power Plant design conforms to GDC 22.

Relevant FSAR Chapters and Sections Chapter 7.1 Fundamental Design Principles 3.4 Criterion 23-Protection System Failure Modes The protection system shall be designed to fail into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.

Implementation in the NuScale Power Plant Design The MPS uses self-diagnoses to detect fatal faults and fail into a safe state. The SSC associated with the MPS are provided with a constant signal to maintain a non-actuated state. Upon loss of signal, the SSC fail into a safe state.

Conformance or Exception The NuScale Power Plant design conforms to GDC 23.

Relevant FSAR Chapters and Sections Section 3.11 Environmental Qualification of Mechanical and Electrical Equipment Section 4.6 Functional Design of Control Rod Drive System Chapter 7 Instrumentation and Controls 3.5 Criterion 24-Separation of Protection and Control Systems The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system.

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Implementation in the NuScale Power Plant Design The MPS incorporates redundancy in multiple areas so that a single failure or removal from service will not prevent safety functions from being accomplished when required.

The MPS has four redundant groups of signal conditioning and trip determination, two divisions of RTS and ESFAS, and redundant communication paths. Each safety function uses two-out-four voting and there are two independent, diverse, and redundant divisions of RTS and ESFAS so that a single failure will not prevent the safety function from being accomplished.

The MPS does not have any connections between divisions. Qualified, safety-related, one way isolation devices are used to send data from the MPS to nonsafety-related systems and to provide input from nonsafety-related systems to the protection systems.

Conformance or Exception The NuScale Power Plant design conforms to GDC 24.

Relevant FSAR Chapters and Sections Chapter 7 Instrumentation and Controls 3.6 Criterion 25-Protection System Requirements for Reactivity Control Malfunctions The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods.

Implementation in the NuScale Power Plant Design The setpoints of the MPS will assure that reactor trip or engineered safety feature actuation occurs before the process reaches the analytical limit. The setpoints are chosen to assure the plant can operate and experience expected operational transients without unnecessary trips or engineered safety feature actuations. Chapter 15 safety analyses demonstrate that the control rod drive system (CRDS) with any assumed credible failure of any single active component is capable of performing a reactor trip when plant parameters exceed the reactor trip setpoint, in accordance with GDC 25.

Conformance or Exception The NuScale Power Plant design conforms to GDC 25.

Relevant FSAR Chapters and Sections Section 4.3 Nuclear Design 2 3.1-19 Revision 4

Chapter 7 Instrumentation and Controls Chapter 15 Transient and Accident Analyses 3.7 Criterion 26-Reactivity Control System Redundancy and Capability Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.

Implementation in the NuScale Power Plant Design The NuScale Power Plant design incorporates two independent reactivity control systems of different design principle: CRDS and the chemical and volume control system (CVCS), in conjunction with the boron addition system.

The CRDS is designed with appropriate margin to assure its reactivity control function under conditions of normal operation, including AOOs. The CRDS facilitates reliable operator control by performing a safe shutdown via gravity-dropping of the control rod assemblies (CRAs) on a reactor trip signal or loss of power. The CRDS is designed such that core reactivity can be safely controlled and that sufficient negative reactivity exists to maintain the core subcritical under cold conditions.

The CVCS operates in conjunction with the boron addition system to satisfy GDC 26 as the second reactivity control system. The CVCS has the ability to control the soluble boron concentration to compensate for fuel depletion during operation and xenon burnout reactivity changes, to assure acceptable fuel design limits are not exceeded.

The CVCS is designed to maintain the reactor as subcritical under cold conditions.

Conformance or Exception The NuScale Power Plant design conforms to GDC 26.

Relevant FSAR Chapters and Sections Section 3.9.4 Control Rod Drive System Section 4.3 Nuclear Design Section 4.6 Functional Design of Control Rod Drive System 2 3.1-20 Revision 4

Chapter 15 Transient and Accident Analyses 3.8 Criterion 27-Combined Reactivity Control Systems Capability The reactivity control systems shall be designed to have a combined capability, in conjunction with poison addition by the emergency core cooling system, of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained.

Implementation in the NuScale Power Plant Design GDC 27 is not applicable to the NuScale design. The following PDC has been adopted:

The reactivity control systems shall be designed to have a combined capability of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods the capability to cool the core is maintained. Following a postulated accident, the control rods shall be capable of holding the reactor core subcritical under cold conditions with all rods fully inserted.

Consistent with GDC 27, this PDC requires that the reactivity control systems function, together with heat removal systems, to protect the core from unacceptable damage under accident conditions. This protection function is met by providing sufficient reactivity control such that core cooling is maintained under accident conditions, analyzed using conservative methodology and assumptions including margin equivalent to the highest worth rod stuck out. Under the NuScale design basis, during normal operation sufficient negative reactivity is maintained (instantaneous shutdown margin) to ensure that the capability to cool the core is maintained under accident conditions by rapid control rod insertion with the highest worth rod stuck out.

The PDC also includes a post-accident holddown criterion specific to the NuScale design. This provision requires the control rods to be capable of maintaining the core subcritical under cold conditions following a postulated accident, without margin for the highest worth rod stuck out. Conservative analysis indicates that a post-accident return to power could occur following initial shutdown, under the condition that the highest worth CRA does not insert. The CVCS system is capable of providing negative reactivity but is not credited in this analysis since it is not a safety-related system.

Section 15.0.6 demonstrates that the passive heat removal safety systems provide sufficient thermal margin such that a return to power does not result in the failure of the fuel cladding fission product barrier, as demonstrated by not exceeding SAFDLs for the analyzed events.

The reactivity control capability required by either GDC 27 or PDC 27 provides assurance that even if a postulated accident damages fuel, continued core cooling will not be precluded and thus accident consequences can be maintained within acceptable limits. The NuScale design assures that fuel cladding integrity is maintained for all design basis events, including postulated accidents, such that the effect of a postulated return to power with failed fuel has not been evaluated in the analysis of 2 3.1-21 Revision 4

postulated accident conditions.

Conformance or Exception The NuScale Power Plant design departs from GDC 27 and supports an exemption from the criterion. The NuScale Power Plant design conforms to PDC 27.

Relevant FSAR Chapters and Sections Section 3.9.4 Control Rod Drive System Section 4.2 Fuel System Design Section 4.3 Nuclear Design Section 4.6 Functional Design of Control Rod Drive System Section 6.3 Emergency Core Cooling System Section 9.3.4 Chemical and Volume Control System Chapter 15 Transient and Accident Analyses 3.9 Criterion 28-Reactivity Limits The reactivity control systems shall be designed with appropriate limits on the potential amount and rate of reactivity increase to assure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core. These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition.

Implementation in the NuScale Power Plant Design The NuScale design places limits on the worth of CRAs, the maximum CRA withdrawal rate, and the CRA insertion. The maximum worth of control rods and control rod insertion limits preclude rupture of the RCPB due to a rod withdrawal or rod ejection accident. Section 15.4 addresses plant safety associated with the reactivity insertion rates.

Conformance or Exception The NuScale Power Plant design conforms to GDC 28.

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Section 4.3 Nuclear Design Section 4.6 Functional Design of Control Rod Drive System Chapter 7 Instrumentation and Controls Section 9.3.4 Chemical and Volume Control System Chapter 15 Transient and Accident Analyses 3.10 Criterion 29-Protection Against Anticipated Operational Occurrences The protection and reactivity control systems shall be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

Implementation in the NuScale Power Plant Design The CRDS and the protection systems are designed to assure a high probability of performing the required safety-related functions in the event of AOO.

The CRDS can perform safety-related functions to control the reactor within fuel and plant limits during AOOs despite a single failure of the system. The CRDS performs a safe shutdown via gravity-dropping of the CRAs on a reactor trip signal or loss of power. The CRDS maintains an ASME BPVC,Section III Division 1, Subsection NB Class 1 boundary for the reactor coolant during normal, upset, emergency, and faulted operating conditions. The safety-related reactor trip function of the CRDS is initiated by MPS through the RTS. The CRDS performs a reactor trip when plant parameters exceed the reactor trip setpoint. Therefore, the reactor is placed in a subcritical condition with any assumed credible failure of any single active component.

The protection systems are designed with sufficient redundancy and diversity to assure high probability of accomplishing their safety-related functions in the event of AOOs.

Conformance or Exception The NuScale Power Plant design conforms to GDC 29.

Relevant FSAR Chapters and Sections Section 3.9.4 Control Rod Drive System Section 4.6 Functional Design of Control Rod Drive System Chapter 7 Instrumentation and Controls Section 9.3.4 Chemical and Volume Control System 2 3.1-23 Revision 4

4 Fluid Systems 4.1 Criterion 30-Quality of Reactor Coolant Pressure Boundary Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical.

Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage.

Implementation in the NuScale Power Plant Design The RPV and pressure retaining components associated with the RCPB are designed, fabricated, and tested in accordance with ASME BPVC,Section III Division 1, Subsection NB, Class 1 are consistent with 10 CFR 50.3 and 10 CFR 50.55a.

The containment evacuation system supports two methods for detecting and, to the extent practical, identifying the source of reactor coolant leakage. These leak detection methods are CNV pressure monitoring and containment evacuation system sample tank level change monitoring. Both leak detection methods are consistent with the guidance in Regulatory Guide 1.45.

Conformance or Exception The NuScale Power Plant design conforms to GDC 30.

Relevant FSAR Chapters and Sections Section 3.2 Classification of Structures, Systems, and Components Section 3.9.6 Functional Design, Qualification and Inservice Testing Program for Pumps, Valves and Dynamic Restraints Section 3.13 Threaded Fasteners (ASME Code Class 1, 2, and 3)

Section 5.2 Integrity of Reactor Coolant Boundary Section 5.3 Reactor Vessel Section 9.3.6 Containment Evacuation System and Containment Flooding and Drain System Section 11.5 Process and Effluent Radiation Monitoring Instrumentation and Sampling 4.2 Criterion 31-Fracture Prevention of Reactor Coolant Pressure Boundary The reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated 2 3.1-24 Revision 4

consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determining (1) material properties, (2) the effects of irradiation on material properties, (3) residual, steady state and transient stresses, and (4) size of flaws.

Implementation in the NuScale Power Plant Design Overpressure protection is provided for the RCPB during low temperature conditions to assure the pressure boundary behaves in a non-brittle manner and the probability for rapidly propagating fracture is minimized. The ferritic materials provide sufficient margin to account for uncertainties associated with flaws and the effects of service and operating conditions.

Conformance or Exception The NuScale Power Plant design conforms to GDC 31.

Relevant FSAR Chapters and Sections Section 3.13 Threaded Fasteners (ASME Code Class 1, 2, and 3)

Section 5.2 Integrity of Reactor Coolant Boundary Section 5.3 Reactor Vessel Section 6.1 Engineered Safety Feature Materials 4.3 Criterion 32-Inspection of Reactor Coolant Pressure Boundary Components which are part of the reactor coolant pressure boundary shall be designed to permit (1) periodic inspection and testing of important areas and features to assess their structural and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel.

Implementation in the NuScale Power Plant Design Components which are part of the RCPB are designed and provided with access to permit periodic inspection and testing requirements for ASME BPVC,Section III Division 1, Subsection NB Class 1 pressure-retaining components in accordance with ASME BPVC,Section XI Division 1 (Reference 3.1-5) pursuant to 10 CFR 50.55a(g). Equipment that may require inspection or repair is placed in an accessible position to minimize time and radiation exposure during refueling and maintenance outages. Plant technicians may access components without being placed at risk for dose or situations where excessive plates, shields, covers, or piping must be moved or removed in order to access components.

The RPV material surveillance program monitors changes in the fracture toughness properties. Specimens are periodically removed and tested in order to monitor 2 3.1-25 Revision 4

ASTM E185-82 (Reference 3.1-6), as required by 10 CFR 50, Appendix H. Table 5.3-2 lists the specimen matrix for the NuScale material surveillance program requirements.

Conformance or Exception The NuScale Power Plant design conforms to GDC 32.

Relevant FSAR Chapters and Sections Section 3.9.6 Functional Design, Qualification and Inservice Testing of Pumps, Valves and Dynamic Restraints Section 5.2.4 Reactor Coolant Pressure Boundary Inservice Inspection and Testing Section 5.3.1 Reactor Vessel Materials 4.4 Criterion 33-Reactor Coolant Makeup A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided. The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system shall be designed to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished using the piping, pumps, and valves used to maintain coolant inventory during normal reactor operation.

Implementation in the NuScale Power Plant Design The CVCS provides reactor coolant makeup during normal operation for small leaks in the RCPB, but is not relied upon during a design basis event. The RPV and CNV design retain sufficient RCS inventory that, in conjunction with safety actuation setpoints to isolate CVCS from the RCS and operation of emergency core cooling system (ECCS),

adequate cooling is maintained and the SAFDLs are not exceeded in the event of a small break in the RCPB.

Conformance or Exception The NuScale Power Plant design does not conform to GDC 33. The NuScale design supports an exemption from the criterion.

Relevant FSAR Chapters and Sections Section 8.2 Offsite Power System Section 8.3 Onsite Power Systems 2 3.1-26 Revision 4

4.5 Criterion 34-Residual Heat Removal A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Implementation in the NuScale Power Plant Design The NuScale design supports an exemption from the power provisions of GDC 34. The following PDC has been adopted:

A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that the system safety function can be accomplished, assuming a single failure.

The decay and residual heat removal safety function is performed by the DHRS flowpath and containment isolation function of the containment system performed by the main steam isolation valves (MSIVs), the main steam isolation bypass valves, and feedwater isolation valves.

The DHRS is a closed-loop, passive condenser design that utilizes circulation flow from the steam generators to dissipate residual and decay core heat to the UHS. The DHRS consists of two independent subsystems, each capable of performing the system safety function in the event of a single failure. The DHRS actuation valves actuate upon loss or an interruption of electrical power.

Conformance or Exception The NuScale Power Plant design conforms to PDC 34.

Relevant FSAR Chapters and Sections Section 5.4.3 Decay Heat Removal System Section 8.2 Offsite Power System 2 3.1-27 Revision 4

Chapter 10 Steam and Power Conversion System Chapter 15 Transient and Accident Analyses 4.6 Criterion 35-Emergency Core Cooling A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Implementation in the NuScale Power Plant Design The NuScale design supports an exemption from the power provisions of GDC 35. The following PDC has been adopted:

A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that the system safety function can be accomplished, assuming a single failure.

The ECCS provides adequate passive heat removal following any loss of reactor coolant event.

The ECCS is fully enclosed inside containment and consists of three reactor vent valves located on the head of the RPV and two reactor recirculation valves located on the side of the RPV. All five valves are closed during normal operation and open when the system is actuated during accident conditions. The reactor vent valves allow steam to flow from the RPV into the CNV, where it then condenses on the CNV walls and collects at the bottom of the CNV. The condensed coolant then reenters the RPV through the reactor recirculation valves and is recirculated to cool the reactor core. The placement of the two reactor recirculation valves assures that the coolant level in the RPV is maintained above the core and the fuel remains covered at all times during ECCS operation.

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single active or passive component failure. The valves are the only active components in the ECCS and are designed to actuate on stored energy. After the actuation, the valves do not require a subsequent change of state or continuous availability of power to maintain their intended safety functions.

Leakage from the RCS to the CNV is detectable by containment pressure instruments, and instrumentation and operation records from the containment evacuation system.

Conformance or Exception The NuScale Power Plant design conforms to PDC 35.

Relevant FSAR Chapters and Sections Section 4.2 Fuel System Design Section 6.3 Emergency Core Cooling System Section 8.2 Offsite Power System Section 8.3 Onsite Power Systems Chapter 15 Transient and Accident Analyses 4.7 Criterion 36-Inspection of Emergency Core Cooling System The emergency core cooling system shall be designed to permit appropriate periodic inspection of important components, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping, to assure the integrity and capability of the system.

Implementation in the NuScale Power Plant Design The ECCS provides accessibility for appropriate periodic inspection of important components in accordance with ASME BPVC,Section III Division 1 to assure the integrity and capability of the system.

Conformance or Exception The NuScale Power Plant design conforms to GDC 36.

Relevant FSAR Chapters and Sections Section 6.3 Emergency Core Cooling System 4.8 Criterion 37-Testing of Emergency Core Cooling System The emergency core cooling system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its 2 3.1-29 Revision 4

to design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.

Implementation in the NuScale Power Plant Design The MPS provides the capability to perform periodic pressure and functional testing of the ECCS that ensures operability and performance of system components and the operability and performance of the system as a whole.

Functional testing of ECCS valves under conditions similar to design conditions is only possible with a differential pressure established between the RPV and the CNV because the main valve control chamber must vent to the CNV. These tests are therefore conducted under conditions that are colder than would exist for a required actuation of the ECCS valves and at a lower differential pressure.

Conformance or Exception The NuScale Power Plant design conforms to GDC 37.

Relevant FSAR Chapters and Sections Section 3.9.6 Functional Design, Qualification and Inservice Testing of Pumps, Valves and Dynamic Restraints Section 6.3 Emergency Core Cooling System 4.9 Criterion 38-Containment Heat Removal A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Implementation in the NuScale Power Plant Design The NuScale design supports an exemption from the power provisions of GDC 38. The following PDC has been adopted:

A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of 2 3.1-30 Revision 4

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that the system safety function can be accomplished, assuming a single failure.

Containment heat removal is an inherent characteristic assured by the materials and physical configuration of the CNV partially immersed in the UHS. The containment heat removal function is accomplished with the passive transfer of containment heat via the steel wall of the NuScale CNV to the UHS. The design configuration of the CNV and UHS provides the ability to remove containment heat rapidly for accident conditions to establish low containment pressure and temperature, and maintain these conditions for an indefinite period with no reliance on active components or electrical power.

During a postulated design basis loss-of-coolant or other conditions involving mass and energy release into containment, the released inventory is collected and accumulates within the CNV. The reactor coolant inventory condenses and accumulates in the CNV. The subsequent actuation of the ECCS establishes a natural circulation coolant pathway that circulates reactor coolant inventory through the CNV volume back to the RPV and through the reactor core.

Conformance or Exception The NuScale Power Plant design conforms to PDC 38.

Relevant FSAR Chapters and Sections Section 6.2.1 Containment Functional Design Section 6.2.2 Containment Heat Removal Section 8.2 Offsite Power System Section 8.3 Onsite Power Systems Section 9.2.5 Ultimate Heat Sink 4.10 Criterion 39-Inspection of Containment Heat Removal System The containment heat removal system shall be designed to permit appropriate periodic inspection of important components, such as the torus, sumps, spray nozzles, and piping to assure the integrity and capability of the system.

Implementation in the NuScale Power Plant Design The major components that provide for the passive containment heat removal function are designed to allow inspections in accordance with in ASME BPVC,Section XI Division 1. The design permits appropriate periodic examination of the CNV to ensure continuing integrity and capability for heat transfer, i.e., the design allows for 2 3.1-31 Revision 4

Conformance or Exception The NuScale Power Plant design conforms to GDC 39.

Relevant FSAR Chapters and Sections Section 6.2.2 Containment Heat Removal 4.11 Criterion 40-Testing of Containment Heat Removal System The containment heat removal system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole, and under conditions as close to the design as practical the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.

Implementation in the NuScale Power Plant Design The NPM passive containment cooling does not include or require active components to provide the containment heat removal function, thus periodic and operation testing specified by GDC 40 does not apply. Testing of the passive containment heat removal function for LOCA conditions was performed and showed that following a design basis event that results in containment pressurization, containment pressure is rapidly reduced and maintained below the design value without operator action. The continuing operability and performance of the containment heat removal function is ensured through periodic inspections, pursuant to GDC 39. Therefore, the underlying intent of GDC 40 is met.

Conformance or Exception The NuScale Power Plant design does not conform to GDC 40. The NuScale design supports an exemption from the criterion.

Relevant FSAR Chapters and Sections Section 3.9.6 Functional Design, Qualification and Inservice Testing of Pumps, Valves and Dynamic Restraints Section 6.2.2 Containment Heat Removal 4.12 Criterion 41-Containment Atmosphere Cleanup Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, 2 3.1-32 Revision 4

and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure.

Implementation in the NuScale Power Plant Design The NuScale design supports an exemption from the power provisions of GDC 41. The following PDC has been adopted:

Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities to assure that its safety function can be accomplished, assuming a single failure.

For the NuScale design, there are no containment atmosphere cleanup systems necessary to ensure containment integrity or to reduce fission product release to the environment following postulated accidents. The CNV in conjunction with the containment isolation system is credited to mitigate the consequences of a design basis accident.

Compliance with GDC 41 is met with the NuScale passive design with respect to hydrogen and oxygen control/cleanup. The CNV can withstand the environmental conditions created by burning of hydrogen during the first 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of design basis and beyond design basis accidents, while maintaining structural integrity and safe shutdown capability.

Natural aerosol removal mechanisms inherent in the containment design deplete elemental iodine and particulates in the containment atmosphere. The limited containment leakage and natural fission product control mechanisms result in offsite doses that are less than regulatory limits.

2 3.1-33 Revision 4

The NuScale design reduces the concentration and quality of fission product release to the environment and ensures CNV integrity is maintained following a postulated design basis accident, thus meeting the intent of PDC 41.

Relevant FSAR Chapters and Sections Section 6.2.5 Combustible Gas Control in the Containment Vessel Section 6.5.3 Fission Product Control Systems Section 8.2 Offsite Power System Section 8.3 Onsite Power Systems 4.13 Criterion 42-Inspection of Containment Atmosphere Cleanup Systems The containment atmosphere cleanup systems shall be designed to permit appropriate periodic inspection of important components, such as filter frames, ducts, and piping to assure the integrity and capability of the systems.

Implementation in the NuScale Power Plant Design The design does not include containment atmosphere cleanup systems which are subject to inspections of GDC 42.

Conformance or Exception The NuScale Power Plant design does not include containment atmosphere cleanup systems which are subject to inspections of GDC 42 and therefore the criterion is not applicable.

Relevant FSAR Chapters and Sections Section 6.5.3 Fission Product Control Systems 4.14 Criterion 43-Testing of Containment Atmosphere Cleanup Systems The containment atmosphere cleanup systems shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and performance of the active components of the systems such as fans, filters, dampers, pumps, and valves and (3) the operability of the systems as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the systems into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of associated systems.

2 3.1-34 Revision 4

The NuScale Power design does not include containment atmosphere cleanup systems which are subject to periodic pressure and functional testing of GDC 43.

Conformance or Exception The NuScale Power Plant design does not include containment atmosphere cleanup systems which are subject to the periodic pressure and functional testing of GDC 43 and therefore the criterion is not applicable.

Relevant FSAR Chapters and Sections Section 6.5.3 Fission Product Control Systems 4.15 Criterion 44-Cooling Water A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Implementation in the NuScale Power Plant Design The NuScale design supports an exemption from the power provisions of GDC 44. The following PDC has been adopted:

A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that the system safety function can be accomplished, assuming a single failure.

The cooling water function is provided by the UHS.

The UHS consists of the reactor pool, refueling pool, and spent fuel pool and functions as a cooling water medium for the decay heat removal heat exchangers, NPMs within the reactor pool, and the stored spent fuel assemblies. The UHS maintains the core temperature at acceptably low levels following any LOCA resulting in the initiation of ECCS. The passive cooling feature provided by the UHS does not include active components and does not rely on electrical power to perform its safety function.

2 3.1-35 Revision 4

normal makeup when a low pool water level is detected.

Conformance or Exception The NuScale Power Plant standard design conforms to PDC 44.

Relevant FSAR Chapters and Sections Section 8.2 Offsite Power System Section 8.3 Onsite Power Systems Section 9.2.5 Ultimate Heat Sink 4.16 Criterion 45-Inspection of Cooling Water System The cooling water system shall be designed to permit appropriate periodic inspection of important components, such as heat exchangers and piping, to assure the integrity and capability of the system.

Implementation in the NuScale Power Plant Design The UHS does not include or require active components to perform its passive cooling function. Leak detection surveillance and level instrumentation are provided to monitor the integrity and capability of the UHS.

Conformance or Exception The NuScale Power Plant design conforms to GDC 45.

Relevant FSAR Chapters and Sections Section 9.2.5 Ultimate Heat Sink 4.17 Criterion 46-Testing of Cooling Water System The cooling water system shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity of its components, (2) the operability and the performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to design as practical, the performance of the full operational sequence that brings the system into operation for reactor shutdown and for loss-of-coolant accidents, including operation of applicable portions of the protection system and the transfer between normal and emergency power sources.

2 3.1-36 Revision 4

The UHS requires no active components to perform the required safety functions. The UHS design permits the inspection of important components, such as the pool water level instrumentation, the pool liner, and the outside surfaces of the containment vessels. These inspections and tests assure the system integrity and capability of the UHS heat removal function.

Conformance or Exception The NuScale Power Plant design conforms to GDC 46.

Relevant FSAR Chapters and Sections Section 9.2.5 Ultimate Heat Sink 5 Reactor Containment 5.1 Criterion 50-Containment Design Basis The reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of the peak conditions, such as energy in steam generators and as required by 50.44 energy from metal-water and other chemical reactions that may result from degradation but not total failure of emergency core cooling functioning, (2) the limited experience and experimental data available for defining accident phenomena and containment responses, and (3) the conservatism of the calculation model and input parameters.

Implementation in the NuScale Power Plant Design The CNV is designed to provide a final barrier against release of fission products while accommodating the calculated pressures and temperatures resulting from any design basis LOCA with sufficient margin such that the design leak rates are not exceeded. The CNV design also takes into consideration the pressures and temperatures associated with combustible gas deflagration. The design includes no internal sub-compartments to eliminate the potential for collection of combustible gases and differential pressures resulting from postulated high-energy pipe breaks within containment.

Conformance or Exception The NuScale Power Plant design conforms to GDC 50.

Relevant FSAR Chapters and Sections Section 3.8.2 Steel Containment 2 3.1-37 Revision 4

Section 8.3 Containment Electrical Penetration Assemblies 5.2 Criterion 51-Fracture Prevention of Containment Pressure Boundary The reactor containment boundary shall be designed with sufficient margin to assure that under operating, maintenance, testing, and postulated accident conditions (1) its ferritic materials behave in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the containment boundary material during operation, maintenance, testing, and postulated accident conditions, and the uncertainties in determining (1) material properties, (2) residual, steady state, and transient stresses, and (3) size of flaws.

Implementation in the NuScale Power Plant Design The design, fabrication, and construction materials for the CNV system includes sufficient margin to provide assurance that the containment pressure boundary will not undergo brittle fracture and the probability of rapidly propagating fracture will be minimized under operating, maintenance, and postulated accident conditions. The ferritic containment pressure boundary materials satisfy the fracture toughness criteria for ASME BPVC Section III Division 1, Class 1 and 2 components.

Conformance or Exception The NuScale Power Plant design conforms to GDC 51.

Relevant FSAR Chapters and Sections Section 6.2.7 Fracture Prevention of Containment Vessel 5.3 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.

Implementation in the NuScale Power Plant Design The CNV design allows testing and inspection, other than as anticipated by GDC 52, to assure CNV leakage integrity.

The CNV design utilizes 10 CFR 50, Appendix J, Type B and C tests to quantify containment leakage, thus assuring that the allowable leakage rate values are not exceeded.

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The NuScale Power Plant design does not conform to GDC 52. The NuScale design supports an exemption from the criterion.

Relevant FSAR Chapters and Sections Section 6.2.6 Containment Leakage Testing 5.4 Criterion 53-Provisions for Containment Testing and Inspection The reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leaktightness of penetrations which have resilient seals and expansion bellows.

Implementation in the NuScale Power Plant Design The CNV is designed to allow for sufficient access for inservice inspection of vessel welds and penetrations, and surveillance testing of containment isolation valves (CIVs) and penetration assemblies pursuant to ASME BPVC,Section XI Division 1 and "Standards and Guides for Operation and Maintenance of Nuclear Power Plants," ASME OM-2012 (Reference 3.1-7).

Conformance or Exception The NuScale Power Plant design conforms to GDC 53.

Relevant FSAR Chapters and Sections Section 3.8.2 Steel Containment Section 6.2.6 Containment Leakage Testing 5.5 Criterion 54-Piping Systems Penetrating Containment Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits.

Implementation in the NuScale Power Plant Design The piping systems that penetrate the CNV are designed with leak detection, isolation, and containment capabilities that are redundant and reliable. The containment isolation components include CIVs and passive containment isolation barriers that are periodically tested to ensure leakage is maintained within acceptable limits. The CIVs close for an ESFAS containment system isolation actuation signal, including when the 2 3.1-39 Revision 4

Conformance or Exception The NuScale Power Plant design conforms to GDC 54.

Relevant FSAR Chapters and Sections Section 3.9.6 Functional Design, Qualification and Inservice Testing of Pumps, Valves and Dynamic Restraints Section 5.2 Integrity of Reactor Coolant Boundary Section 5.4 Reactor Coolant System Component and Subsystem Design Section 6.2 Containment Systems 5.6 Criterion 55-Reactor Coolant Pressure Boundary Penetrating Containment Each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or

2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or

3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or

4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

Isolation valves outside containment shall be located as close to containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.

Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or of lines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomena, and additional isolation valves and containment, shall include 2 3.1-40 Revision 4

Implementation in the NuScale Power Plant Design The lines that are part of the RCPB and penetrate primary reactor containment are designed to provide adequate containment isolation. The RCS injection line, pressurizer spray supply line, and RCS discharge line, in addition to the reactor high point degasification line, are part of the RCPB and penetrate primary reactor containment. Consistent with GDC 55 except for the location of the isolation valves, two CIVs are provided for each of these lines and are located outside the CNV. Each line features a single-body, dual valve welded directly to a CNV top head nozzle safe-end to provide two containment isolation barriers in series. The isolation valves are Seismic Category 1 components and constructed in accordance with ASME BPVC,Section III, Division 1, Subsection NB.

Conformance or Exception The NuScale Power design departs from GDC 55. The NuScale design supports an exemption for the lines that depart from the four alternatives for containment isolation valves specified in the criterion.

Relevant FSAR Chapters and Sections Section 6.2.4 Containment Isolation System Chapter 15 Transient and Accident Analyses 5.7 Criterion 56-Primary Containment Isolation Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class of lines, such as instrument lines, are acceptable on some other defined basis:

1) One locked closed isolation valve inside and one locked closed isolation valve outside containment; or

2) One automatic isolation valve inside and one locked closed isolation valve outside containment; or

3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment; or

4) One automatic isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic isolation valve outside containment.

2 3.1-41 Revision 4

designed to take the position that provides greater safety.

Implementation in the NuScale Power Plant Design The lines that connect directly to the containment atmosphere and penetrate primary reactor containment are designed to provide adequate containment isolation. The containment evacuation line and the containment flood and drain line connect directly to the containment atmosphere and penetrate primary reactor containment. The control rod drive closed loop cooling system supply and return lines penetrate primary reactor containment and are conservatively treated as if the lines connect directly to containment atmosphere. Consistent with GDC 56 except for the location of the isolation valves, two CIVs are provided for each of the lines and are located outside the CNV. The lines feature a single-body, dual valve welded directly to a containment top head nozzle safe-ends to provide two containment isolation barriers in series. The isolation valves are Seismic Category 1 components and constructed in accordance with ASME BPVC Section III Division 1, Subsection NB.

Conformance or Exception The NuScale Power design departs from GDC 56. An exemption is provided for the lines that depart from the four alternatives for containment isolation valves specified in the criterion.

Relevant FSAR Chapters and Sections Section 6.2.4 Containment Isolation System 5.8 Criterion 57-Closed System Isolation Valves Each line that penetrates primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one containment isolation valve which shall be either automatic, or locked closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.

Implementation in the NuScale Power Plant Design The lines that penetrate primary reactor containment and are neither part of the RCPB nor connected directly to the containment atmosphere are designed to provide adequate containment isolation. At least one CIV is provided for each of these lines, with exception of DHRS.

The CIV provided for each applicable main steam and feedwater line is a Seismic Category 1, ASME BPVC,Section III Division 1, Subsection NC, Class 2 valve. As noted in Section 3.1.5.7, for the RCCW return and supply lines, two CIVs are provided for each line in a single-body, dual valve. These valves are Seismic Category 1, ASME BPVC,Section III Division 1, Subsection NB, Class 1 components.

2 3.1-42 Revision 4

outside containment and does not have CIVs. Two isolation barriers are provided by the direct connection of the closed-loop DHRS outside containment, and by the closed-loop inside of containment formed by the steam generator system within the RPV, and the connecting piping. The DHRS is a welded Seismic Category I, ASME BPVC,Section III Division 1, Subsection NC, Class 2 design with a design temperature and pressure rating equal to that of the RPV and meets the applicable criteria of NRC Branch Technical Position 3-4, Revision 2.

Conformance or Exception The NuScale Power Plant design departs from GDC 57. The NuScale design supports an exemption for the lines that depart from the isolation barriers specified in the criterion.

Relevant FSAR Chapters and Sections Section 5.4.3 Decay Heat Removal System Section 6.2.4 Containment Isolation System 6 Fuel and Radioactivity Control 6.1 Criterion 60-Control of Releases of Radioactive Materials to the Environment The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such effluents to the environment.

Implementation in the NuScale Power Plant Design The NuScale Power Plant is designed to control and minimize the release of radioactive materials in solid waste and gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation and AOOs. Alarm setpoints, design features, and automated isolation features ensure compliance with GDC 60 and that the limitations of 10 CFR 20 and 10 CFR 50, Appendix I are not exceeded.

Conformance or Exception The NuScale Power Plant design conforms to GDC 60.

Relevant FSAR Chapters and Sections Section 9.1.3 Spent Fuel Pool Cooling and Cleanup System Section 9.2 Water Systems 2 3.1-43 Revision 4

Chapter 11 Radioactive Waste Management Chapter 15 Transient and Accident Analyses 6.2 Criterion 61-Fuel Storage and Handling and Radioactivity Control The fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

Implementation in the NuScale Power Plant Design The spent fuel pool cooling system cools the spent fuel assemblies stored in the fuel storage racks in the spent fuel pool for normal operating conditions. Water in the spent fuel pool shields the assemblies and normal makeup for evaporation is provided by the demineralized water system. The UHS performs the cooling and shielding functions under accident conditions. The pool cleanup system purifies the shared body of water in the spent fuel pool, the reactor pool, and the refueling pool that make up the UHS.

This system has filters and demineralizers for pool water cleanup, and provisions for periodic sampling.

The large inventory of water in the UHS is a passive source of water that ensures the water level in the spent fuel pool remains above the stored spent fuel assemblies for weeks without additional makeup water to the UHS and without operation of the two active cooling systems. Section 9.2.5 describes performance of the UHS for accident conditions.

The area around the spent fuel pool is serviced by nonsafety-related Reactor Building heating and ventilation system, which controls the release of airborne radionuclides from evaporating UHS pool water for normal operating conditions. For accident conditions, the radiological consequences of a fuel handling accident are addressed in Chapter 15.

The piping penetrations through the walls of the UHS pool and the piping in the pool can not drain the water and adversely affect the inventory of water available for cooling and shielding the spent fuel assemblies.

The design of the spent fuel storage facility, the active pool cooling and cleanup systems, and the UHS satisfy GDC 61.

Permanent plant shielding is described in Section 12.3 and radiation monitoring is described in Section 11.5 and Section 12.3.

2 3.1-44 Revision 4

Conformance or Exception The NuScale Power Plant design conforms to GDC 61.

Relevant FSAR Chapters and Sections Section 9.1 Fuel Storage and Handling Section 9.2.5 Ultimate Heat Sink Section 9.3.4 Chemical and Volume Control System Section 9.4.2 Reactor Building and Spent Fuel Pool Area Ventilation System Chapter 11 Radioactive Waste Management Chapter 12 Radiation Protection Chapter 15 Transient and Accident Analysis 6.3 Criterion 62-Prevention of Criticality in Fuel Storage and Handling Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

Implementation in the NuScale Power Plant Design The design and controls for operation of the fuel handling equipment and fuel storage racks prevent an inadvertent criticality by use of geometrically safe configurations, as well as plant programs and procedures. Section 9.1 describes criticality safety for handling and storage of new and spent fuel assemblies.

Conformance or Exception The NuScale Power Plant design conforms to GDC 62.

Relevant FSAR Chapters and Sections Section 9.1 Fuel Storage and Handling 6.4 Criterion 63-Monitoring Fuel and Waste Storage Appropriate systems shall be provided in fuel storage and radioactive waste systems and associated handling areas (1) to detect conditions that may result in loss of residual heat removal capability and excessive radiation levels and (2) to initiate appropriate safety actions.

2 3.1-45 Revision 4

Monitoring for the loss of decay heat removal capability and excessive radiation levels is provided in the fuel storage and radioactive waste systems and associated handling areas for both normal and accident conditions. Information on cooling system performance is provided by the temperature detectors on the inlets and outlets of the heat exchangers in the spent fuel pool cooling system and reactor pool cooling system.

The outlet temperature detectors have a high set point for an alarm that alerts operators to determine the cause and ensure adequate active cooling performance.

Leakage from the liner in the UHS pools is collected by the pool leakage detection system and directed to sumps in the radioactive waste drain system for detection.

Leakage from the piping and equipment in the pool cooling and cleanup systems is also collected by sumps in the radioactive waste drain system for detection. For normal and accident conditions, the UHS system provides redundant pool water level instruments. Radiation monitoring equipment is provided to detect excessive radiation levels and initiate appropriate alarms and procedural actions.

Conformance or Exception The NuScale Power Plant design conforms to GDC 63.

Relevant FSAR Chapters and Sections Section 9.1.2 New and Spent Fuel Storage Section 9.1.3 Spent Fuel Pool Cooling and Cleanup System Section 9.3.2 Process Sampling System Section 9.4.2 Reactor Building and Spent Fuel Pool Area Ventilation System Section 11.5 Process and Effluent Radiation Monitoring Instrumentation and Sampling Chapter 12 Radiation Protection 6.5 Criterion 64-Monitoring Radioactivity Releases Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

Implementation in the NuScale Power Plant Design The NuScale Power Plant provides means to monitor gaseous and liquid radioactivity releases resulting from normal operation, including AOOs, and from postulated accidents.

2 3.1-46 Revision 4

discharge paths and in the plant environs are monitored during normal and accident conditions by the radiation monitors.

Area radiation monitors supplement the personnel and area radiation survey provisions of the radiation protection program described in Section 12.5. Process and effluent radiation monitors provide alarm, indication, and archiving features to the main control room. These monitors provide the ability to measure and record the release of radioactive liquids and gases via the effluent release paths and into the plant environs.

Measurement capability and reporting of effluents are based on the guidelines of Regulatory Guides 1.183 and 1.21.

Conformance or Exception The NuScale Power Plant design conforms to GDC 64.

Relevant FSAR Chapters and Sections Section 9.1.3 Spent Fuel Pool Cooling and Cleanup System Section 9.2.2 Reactor Component Cooling Water System Section 9.2.9 Utility Water Systems Section 9.3 Process Auxiliaries Section 9.4.2 Reactor Building and Spent Fuel Pool Area Ventilation System Chapter 11 Radioactive Waste Management Chapter 12 Radiation Protection 7 References 3.1-1 American Society of Mechanical Engineers, Quality Assurance Requirements for Nuclear Facility Applications, ASME NQA-1-2008/1a-2009 Addenda, New York, NY.

3.1-2 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, "Class 1 Components," 2013 edition,Section III Division 1, Subsection NB, New York, NY.

3.1-3 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, "Materials," 2013 edition,Section II, New York, NY.

2 3.1-47 Revision 4

York, NY.

3.1-5 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, "Rules for Inservice Inspection of Nuclear Components," 2013 edition,Section XI Division 1, New York, NY.

3.1-6 American Society for Testing and Materials, Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels, ASTM E185-1982, Philadelphia, PA.

3.1-7 American Society of Mechanical Engineers, "Standards and Guides for Operation and Maintenance of Nuclear Power Plants," ASME OM-2012, New York, NY.

2 3.1-48 Revision 4

Structures, systems, and components (SSC) are classified according to nuclear safety classification, seismic category, and quality group. This classification aids the determination of the appropriate quality standards and the identification of applicable codes and standards. SSC classification is based on a consideration of both safety-related function (consistent with the definition of safety related in 10 CFR 50.2) and risk significant functions determined as part of the design reliability assurance program. The design reliability assurance program process is described in Section 17.4.

SSC are classified as A1, A2, B1, and B2 in accordance with their safety and risk categories:

A1 - SSC that are determined to be both safety-related and risk-significant

A2 - SSC that are determined to be both safety-related and not risk-significant

B1 - SSC that are determined to be both nonsafety-related and risk-significant

B2 - SSC that are determined to be both nonsafety-related and not risk-significant Certain nonsafety-related SSC that perform risk-significant functions require regulatory oversight. The required oversight is identified by the regulatory treatment of nonsafety systems (RTNSS) process as discussed in Section 19.3.

Table 3.2-1 provides the listing of SSC, including designation of classification, seismic category, and quality group. For the listed SSC, Table 3.2-1 also identifies applicable augmented design requirements and the applicable quality assurance program requirements. The systems are listed in Table 3.2-1 alpha-numerically by system codes. Within a given system, the SSC are listed, generally, in the order of the SSC classification (i.e., A1, A2, B1, and B2). Structures that are of conceptual design are listed within double brackets in Table 3.2-1.

Seismic and quality group classification is described in Section 3.2.1 and Section 3.2.2, respectively.

The SSC classification process is applied at the component level based upon the system functions performed. At the system level, system functions are designated as safety-related or nonsafety-related, and risk-significant or not risk-significant. Components are then classified commensurate with the safety and risk-significance of the system function(s) they support. A system that primarily performs safety-related or risk-significant functions may include nonsafety-related, not risk-significant components, on the basis of those components only supporting nonsafety-related, not risk-significant secondary system functions. Similarly, components that support multiple system functions may include multiple design features, each related to the different system functions. Components with any safety or risk design feature are classified on the basis of that feature.

Safety-related SSC and risk-significant SSC are subject to the Quality Assurance program requirements described in Section 17.5 and documented in the applicable quality assurance program column of Table 3.2-1. In addition, all or part of 10 CFR 50 Appendix B has been applied to some nonsafety-related SSC where specific regulatory guidance applies (e.g.,

Regulatory Guide (RG) 1.29). The application of 10 CFR 50, Appendix B to specific nonsafety-related SSC is included in Table 3.2-1.

2 3.2-1 Revision 4

nonsafety-related (based on the definition in 10 CFR 50.2). The selection of augmented requirements is based on a consideration of the important functionality to be performed by the nonsafety-related SSC and regulatory guidance applicable to the functionality (e.g., consistent with the functionality specified in General Design Criterion 60 for controlling radioactive effluents, augmented requirements are specified for radwaste systems based on the guidance in RG 1.143). Augmented design requirements, if applicable, are identified in Table 3.2-1.

The principal codes and standards used for the design of safety-related and risk-significant SSC are in accordance with the guidance of Regulatory Guide (RG) 1.26. If additional standards are invoked, they are noted in Table 3.2-1.

Item 3.2-1: A COL applicant that references the NuScale Power Plant design certification will update Table 3.2-1 to identify the classification of site-specific structures, systems, and components.

1 Seismic Classification Seismic classification of SSC is consistent with the guidance of RG 1.29, Seismic Design Classification for Nuclear Power Plants, Revision 5, with the following exception. SSC that meet Staff Regulatory Guidance C.1.i are designated Seismic Category II rather than Seismic Category I consistent with industry precedent and practice. Seismic classification uses the following categories: Seismic Category I, Seismic Category II, Seismic Category III, and Seismic Category RW-IIa. These categories are described in Section 3.2.1.1, Section 3.2.1.2, Section 3.2.1.3, and Section 3.2.1.4, respectively.

Some nonsafety-related SSC are designated Seismic Category I as an augmenting requirement if the function is required following an earthquake.

In addition to RG 1.29, seismic categorization of SSC is also consistent with the guidance in RG 1.143 "Design Guidance For Radioactive Waste Management Systems, Structures, And Components Installed In Light-Water-Cooled Nuclear Power Plants"; and RG 1.189 "Fire Protection For Nuclear Power Plants."

RG 1.143 establishes design criteria for three different levels of radioactive waste content.

The application of RG 1.143 with respect to radioactive waste management systems is discussed in Sections 11.2, 11.3 and 11.4. Seismic design expectations for radioactive waste management SSC are discussed in Section 3.2.1.4.

The seismic classification of instrumentation sensing lines is in accordance with RG 1.151, as discussed in Section 7.2.2 and in Section C.1.f of RG 1.29. The use of this guidance assures that the instrument sensing lines used to actuate or monitor safety-related functionality are appropriately classified as Seismic Category I and are capable of withstanding the effects of the SSE.

The design of fire protection systems in accordance with RG 1.189 is described in Section 9.5.1, and its classification is included in Table 3.2-1.

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SSC classified as safety-related are designed to be capable of performing their safety functions during and following a safe shutdown earthquake (SSE). Therefore, these safety-related SSC, including their foundations and supports, are classified as Seismic Category I.

Some SSC classified as nonsafety-related are also designed to be capable of performing their nonsafety-related functions during and following an SSE. These nonsafety-related SSC, including their foundations and supports, are also classified as Seismic Category I.

Seismic Category I SSC are designed to withstand the seismic loads associated with the SSE, in combination with other designated loads, without loss of function or pressure integrity. Development of SSE seismic design loads is addressed in Section 3.7. The design of Seismic Category I structures is addressed in Section 3.8. The seismic design of mechanical systems and components is addressed in Section 3.9. The seismic qualification of mechanical and electrical equipment, including their supports, is addressed in Section 3.10.

Use of Seismic Category I piping is minimized in the NuScale Power Plant design. Drain lines, vent lines, fill lines, and test lines coming off the Seismic Category I piping are treated as part of the Seismic Category I piping.

For systems that are partially Seismic Category I, the Category I portion of the system extends to the first seismic restraint beyond the isolation valves that isolate the part that is Seismic Category I from the non-seismic portion of the system.

At the interface between Seismic Category I and non-seismic systems, the Seismic Category I dynamic analysis requirements are extended to either the first anchor point in the non-seismic system or a sufficient distance into the non-Seismic Category I system so that the Seismic Category I analysis remains valid.

Safety-related and nonsafety-related, Seismic Category I SSC are subject to the pertinent quality assurance program requirements of 10 CFR 50, Appendix B.

1.2 Seismic Category II The design requirements in Staff Regulatory Guidance C.1.i in RG 1.29 for protection of Seismic Category I SSC are applied as follows to SSC classified as Seismic Category II.

SSC that perform no safety-related function, but whose structural failure or adverse interaction could degrade the functioning or integrity of a Seismic Category I SSC to an unacceptable level or could result in incapacitating injury to occupants of the control room during or following an SSE, are designed and constructed so that the SSE would not cause such failure. These SSC are classified as Seismic Category II.

Because they are not required to remain functional, the Seismic Category II classification is applied only to the portions of systems where a potential for adverse interaction with a Seismic Category I SSC exists. Additionally, nonsafety-related instrument lines from safety related pressure boundaries are required to maintain pressure integrity.

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1.3 Seismic Category III SSC not classified as Seismic Category I or Seismic Category II are classified as Seismic Category III. This category includes SSC that have no seismic design requirements and SSC that may be subject to seismic design criteria that are incorporated in, or invoked by, an applicable commercial or industry code.

1.4 Safety Classification RW-IIa RG 1.143 establishes design criteria for SSC that contain radioactive waste. Within RG 1.143 SSC are grouped based upon the quantity of radioactive material. Specifically, RG 1.143 uses three classifications: RW-IIa, RW-IIb, and RW-IIc. These design criteria are applied in addition to the seismic categorization. Therefore a SSC that is used for radioactive waste must satisfy both criteria. There are no Seismic Category I SSC that have RG 1.143 design requirements. There is one Seismic Category II SSC that does. The Radioactive Waste Building is Seismic Category II due to its proximity to the Reactor Building, and it is RW-IIa due to its design radioactive material content.

RG 1.143 specifies that RW-IIa SSC are designed to withstand 1/2 of the SSE. As such, the Radioactive Waste Building is designed to both remain intact (satisfying Seismic Category II) when subjected to a full SSE; and intact and functional (satisfying RW-IIa) when subjected to an earthquake with half the force of the SSE.

All other radioactive waste SSC are sufficiently separated from Seismic Category I SSC that they are Seismic Category III.

RG 1.143 classification is included in Table 3.2-1 within the Quality Class column. SSC that are classified as RW-IIb and RW-IIc are designed to industry codes and standards, which conforms with Seismic Category III.

2 System Quality Group Classification Quality group A through D classifications of relevant SSC are performed in accordance with the applicable guidance of RG 1.26 and RG 1.143. Refer to Table 3.2-1 for a listing of the identified classifications.

The quality group boundaries are included on piping and instrument drawings as the third character (Code Identifier) in the Piping Line Class Specification Convention. Code Identifiers A - C correspond to ASME Class 1 through 3 and align with quality groups A - C.

Code identifier D corresponds to Quality Group D as described in RG 1.26.

Safety-related instrument sensing lines are designed and constructed in accordance with ANSI/ISA-67.02.01-1999 (Reference 3.2-2) as described in RG 1.151. The standard ANSI/ISA-67.02.01-1999 establishes the applicable code requirements and code boundaries for the design and installation of instrument sensing lines interconnecting safety-related piping and vessels with both safety-related and nonsafety-related instrumentation. This is further discussed in Section 7.2.2.

2 3.2-4 Revision 4

and steam generator supports and tube supports (see Section 5.4.1.5) comply with the design and construction requirements of Subsection NG of Section III, Division 1 of the ASME BPVC (Reference 3.2-1).

2.1 Quality Group A Quality Group A applies to pressure-retaining components that form part of the reactor coolant pressure boundary, except those that can be isolated from the reactor coolant system by two automatically-closed or normally-closed valves in series.

Quality Group A SSC meet the requirements for Class 1 components in Section III, Division 1 of the ASME BPVC (Reference 3.2-1) and applicable conditions promulgated in 10 CFR 50.55a(b). Supports for Quality Group A SSC meet the requirements for Class 1 supports in Section III, Division 1, Subsection NF of the ASME BPVC and are not separately listed in Table 3.2-1. Exceptions exist for supports within the pressure retaining boundary of the RPV. See Section 3.2.2 and Section 5.4.1.5 for additional information.

The remaining portions of the reactor coolant pressure boundary are in Quality Group B.

2.2 Quality Group B Quality Group B applies to water- and steam-containing pressure vessels, heat exchangers (other than turbines and condensers), storage tanks, piping, pumps, and valves that are:

part of the reactor coolant pressure boundary but are excluded from Quality Group A.

safety-related or risk-significant systems or portions of systems that are designed for (i) emergency core cooling, (ii) post-accident containment heat removal, or (iii) post-accident fission product removal.

safety-related or risk-significant systems or portions of systems that are designed for (i) reactor shutdown or (ii) residual heat removal.

portions of the steam and feedwater systems extending from and including the secondary side of steam generators up to and including the outermost containment isolation valves, and connected piping up to and including the first valve (including a safety or relief valve) that is either normally closed or capable of automatic closure during all modes of normal reactor operation.

systems or portions of systems connected to the reactor coolant pressure boundary that cannot be isolated from that boundary during all modes of operation by two normally closed or automatically closable valves.

Quality Group B SSC meet the requirements for Class 2 components in Section III, Division 1 of the ASME BPVC and applicable conditions promulgated in 10 CFR 50.55a(b). Supports for Quality Group B SSC meet the requirements for Class 2 2 3.2-5 Revision 4

2.3 Quality Group C Quality Group C applies to water-, steam-, and radioactive-waste-containing pressure vessels; heat exchangers (other than turbines and condensers); storage tanks; piping; pumps; and valves that are not part of the reactor coolant pressure boundary or included in Quality Group B but part of the following:

safety-related or risk-significant portions of cooling water and auxiliary feedwater systems that are designed for (i) emergency core cooling, (ii) postaccident containment heat removal, (iii) postaccident containment atmosphere cleanup, or (iv) residual heat removal from the reactor and spent fuel storage pool that (i) do not operate during any mode of normal reactor operation and (ii) cannot be tested adequately

safety-related or risk-significant portions of cooling water and seal water systems that are designed to support the functioning of other safety-related or risk-significant systems and components

portions of systems that are connected to the reactor coolant pressure boundary and capable of being isolated from that boundary by two valves during all modes of normal reactor operation

systems other than radioactive waste management systems that may contain radioactive material and whose postulated failure would result in conservatively calculated potential off-site doses that exceed 0.5 rem to the whole body or its equivalent to any part of the body Quality Group C SSC meet the requirements for Class 3 components in Section III, Division 1 of the ASME BPVC and applicable conditions promulgated in 10 CFR 50.55a(b). Supports for Quality Group C SSC meet the requirements for Class 3 supports in Section III, Division 1, Subsection NF of the ASME BPVC and are not separately listed in Table 3.2-1.

2.4 Quality Group D Quality Group D applies to water and steam-containing components that are not part of the reactor coolant pressure boundary or included in Quality Groups B or C, but are part of systems or portions of systems that contain or may contain radioactive material (and are not radioactive waste management systems).

SSC determined to be Quality Group D in accordance with guidance of RG 1.26 are listed in Table 3.2-1. SSC designated as Quality Group D meet the codes and standards for components identified as applicable for Quality Group D in Table 1 of RG 1.26.

Codes and standards for Quality Group D SSC and their supports are as follows:

Pressure Vessels - ASME BPVC,Section VIII (Reference 3.2-3)

Piping and Valves - ASME B31.1, Power Piping (Reference 3.2-4)

Pumps - Manufacturers standards 2 3.2-6 Revision 4

0-15 psig Storage Tanks - API-620 (Reference 3.2-7) 3 References 3.2-1 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Rules for Construction of Nuclear Facility Components, 2013 edition,Section III, New York, NY.

3.2-2 American National Standards Institute/International Society of Automation Nuclear Safety-Related Instrument-Sensing Line Piping and Tubing Standard for Use in Nuclear Power Plants, ANSI/ISA 67.02.01-1999, Research Triangle Park, NC.

3.2-3 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Rules for Construction of Pressure Vessels,Section VIII, Division 1, New York, NY.

3.2-4 American Society of Mechanical Engineers, Power Piping, ASME B31.1, New York, NY.

3.2-5 American Petroleum Institute, Welded Steel Tanks for Oil Storage, API 650, 12th edition, 2013, Washington, DC.

3.2-6 American Water Works Association, Welded Steel Tanks for Water Storage, AWWA D-100, Denver, Colorado.

3.2-7 American Petroleum Institute, Design and Construction of Large, Welded, Low-pressure Storage Tanks, API 620, 12th edition, 2014, Washington, DC.

2 3.2-7 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

CNTS, Containment System All components (except as listed below) RXB A1 N/A Q None B I

CVC Injection Check Valve RXB B2 None AQ-S None C I

CVC Discharge Excess Flow Check Valve

CVC PZR Spray Check Valve

CVC Injection & Discharge Nozzles RXB A1 N/A Q None A I

CVC PZR Spray Nozzle

CVC PZR Spray CIV

CVC RPV High Point Degasification Nozzle

CVC RPV High Point Degasification CIV

RVV & RRV Trip/Reset # 1 & 2 Nozzles

RVV Trip 1 & 2/Reset #3 Nozzles

CVC Injection & Discharge CIVs

NPM Lifting Lugs RXB B1 None AQ-S

ANSI N14.6 N/A I

Top Support Structure

NUREG-0612

Top Support Structure Diagonal Lifting Braces

CNV Fasteners RXB A1 N/A Q None N/A I

Hydraulic skid

CNV Seismic Shear Lug

CNV CRDM Support Frame

Containment Pressure Transducer (Narrow Range)

Containment Water Level Sensors (Radar Transceiver)

SG 1 & 2 Steam Temperature Sensors (RTD)

CNTS CFDS Piping in containment RXB B2 None AQ-S None B II Piping from (CES, CFDS, FWS, MSS, and RCCWS) CIVs to disconnect flange (outside containment) RXB B2 None AQ-S None D I CVCS Piping from CIVs to disconnect flange (outside containment) RXB B2 None AQ-S None C I CIV Close and Open Position Sensors: RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

CES, Inboard and Outboard

CFDS, Inboard and Outboard

CVCS, Inboard and Outboard PZR Spray Line

CVCS, Inboard and Outboard RCS Discharge

CVCS, Inboard and Outboard RCS Injection

CVCS, Inboard and Outboard RPV High-Point Degasification

RCCWS, Inboard and Outboard Return and Supply

SGS, Steam Supply CIV/MSIVs and CIV/MSIV Bypasses CIV Close and Open Position Indication RXB A1 None Q None N/A I

FWS, Supply to SGs and DHR HXs FWIV Containment Pressure Transducer (Wide Range) RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

Containment Air Temperature (RTDs) RXB B2 None AQ-S None N/A II

FW Temperature Transducers SGS, Steam Generator System

SG tubes RXB A1 N/A Q None A I

Integral steam plenum

Integral steam plenum caps

Feed plenum access ports

SG tube supports RXB A1 N/A Q None N/A I

Upper and lower SG supports Tier 2 3.2-8 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

Steam piping inside containment RXB A2 N/A Q None B I

Feedwater piping inside containment

Feedwater supply nozzles

Main steam supply nozzles

Thermal relief valves

Feed plenum access port covers

Steam plenum access ports

Steam plenum access port covers Flow restrictors RXB A2 N/A Q None N/A I RXC, Reactor Core System Fuel assembly (RXF) RXB A1 N/A Q None N/A I Fuel Assembly Guide Tube RXB A2 N/A Q None N/A I Incore Instrument Tube RXB B2 None AQ-S None N/A I CRDS, Control Rod Drive System

Control Rod Drive Shafts RXB A1 N/A Q None N/A I

Control Rod Drive Latch Mechanism CRDM Pressure Boundary (Latch Housing, Rod Travel Housing, Rod Travel Housing Plug) RXB A2 N/A Q None A I CRDS Cooling Water Piping and Pressure Relief Valve RXB B2 None AQ-S None B II Rod Position Indication (RPI) Coils RXB B2 None AQ-S None N/A I

Control Rod Drive Coils RXB B2 None AQ-S None N/A II

CRDM power cables from EDN breaker to MPS breaker

CRDM power cables from MPS breaker to CRDM Cabinets

CRDM Control Cabinet RXB B2 None AQ None N/A III

CRDM Power & Rod Position Indication Cables

Rod Position Indication Cabinets (Train A/B)

CRA, Control Rod Assembly All components RXB A2 N/A Q None N/A I NSA, Neutron Source Assembly All components RXB B2 None AQ-S None N/A I RCS, Reactor Coolant System All components (except as listed below) RXB A1 N/A Q None A I

Reactor vessel internals (upper riser assembly (Note 7), lower riser assembly, core support assembly, flow RXB A1 None Q None N/A I diverter, and pressurizer spray nozzles)

Reactor vessel internals upper riser bellows-lateral seismic restraining structure RXB A1 N/A Q None N/A I

Reactor vessel internals upper riser bellows-vertical expansion structure RXB B2 N/A AQ-S ASME BPVC Section III Division 1 NG guidance N/A II

Narrow Range Pressurizer Pressure Elements

PZR/RPV Level Elements

Narrow Range RCS Hot Leg Temperature Elements

Wide Range RCS Hot Leg Temperature Elements

RCS Flow Transmitters (Ultrasonic)

Wide Range RCS Pressure Elements RXB A2 N/A Q None N/A I

Wide Range RCS Cold Leg Temperature Elements Reactor Safety Valve Position Indicator RXB B2 None AQ-S Environmental Qualification N/A I Power from EDS

PZR Control Cabinet RXB B2 None AQ-S None N/A II

PZR Vapor Temperature Element

PZR heater power cabling from MPS breaker to PZR heaters

Pressurizer Liquid Temperature Element

Narrow Range RCS Cold Leg Temperature Element PZR heater power cabling from ELV breaker to MPS breaker RXB B2 None None None N/A III Tier 2 3.2-9 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

CVCS, Chemical and Volume Control System DWS Supply Isolation Valves RXB A2 N/A Q None C I Position Indication for DWS Supply Isolation Valves RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

Discharge Spoolpiece Drain Valve RXB B2 None AQ-S None C I

Discharge Spoolpiece Isolation Valve

Injection Spoolpiece Drain Valve

Pressurizer Spoolpiece Drain Valve

NuScale Power Module Removable Spoolpieces

RPV High Point Degasification Isolation Valve

RPV High Point Degasification Spoolpiece Drain Valve Hydrogen bottle and distribution assembly including excess flow valve RXB B2 None AQ-S None D II Pressure Indicating Transmitter for Hydrogen Injection Bottle RXB B2 None AQ-S None N/A II

Mass Flow Instruments for CVC Injection Line RXB B2 None AQ None N/A III

CVC Discharge Line,

CVC Makeup Line,

LRW Letdown Line (Pressure, Temperature, Flow)

Other Instrumentation (Pressure, Temperature, Flow, Radioactivity, Boron) RXB B2 None None None N/A III All other components RXB B2 None None None D III BAS, Boron Addition System All components (except as listed below) RXB B2 None None None D III

Instrumentation (Pressure, Temperature, Flow, Level, Position) RXB B2 None None None N/A III

Hopper Scale

Batch Tank Mixer MHS, Module Heatup System All components (except as listed below) RXB B2 None None None D III

Instrumentation (Pressure, Temperature, Level) RXB B2 None None None N/A III ECCS, Emergency Core Cooling System

Reactor Vent Valve (RVV) RXB A1 N/A Q None A I

RVV Trip Valve

Reactor Recirculation Valve (RRV)

RRV Trip Valve

Reset Valve

Hydraulic lines

RRV Position Indication RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

RVV Position Indication

Trip Valve Position Indication Reset Valve Position Indication RXB B2 None AQ-S None N/A II DHRS, Decay Heat Removal System SG Steam Pressure Instrumentation (4 per side) RXB A1 N/A Q None N/A I

Actuation Valve (2 per side) RXB A2 N/A Q None B I

Condenser (1 per side)

Condenser Outlet Pressure Instrumentation (3 per side) RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

Condenser Outlet Temperature Instrumentation (2 per side)

Valve Position Indicator (2 for open, 2 for close per side)

Level Instrument (2 per side) RXB B2 None AQ-S None N/A II CRHS, Control Room Habitability System All components (except as listed below) CRB B2 None AQ-S None N/A I

Air Supply Isolation Solenoid Valve Position Indicators CRB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

CRE Pressure Relief Isolation Valve Position Indicators Tier 2 3.2-10 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

CRE Differential Pressure Transmitters CRB B2 None AQ-S None N/A II

CRH Bottle Pressure Instruments

Flow Transmitters

Pressure Reducing Valve Pressure Indicators Air compressor and dryer CRB B2 None None None N/A III CRVS, Normal Control Room HVAC All components (except as listed below) CRB B2 None None None N/A III CRE Isolation Damper Position CRB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

CRE Isolation Dampers CRB B2 None AQ-S None N/A I

Fire and Smoke Dampers supporting the MCR

Radiation Monitors (Downstream of charcoal filter unit)

Outside Air intake Smoke Detectors CRB B2 None AQ-S None N/A I

Outside air Isolation Damper Position CRB B2 None AQ-S RG 1.78 N/A I

Toxic gas detectors Outside Air Isolation Dampers for CRV Recirculation Mode CRB B2 None AQ-S

RG 1.78 N/A II

RG 1.140

Backup diesel powered

Charcoal and HEPA filtered

Maintain Positive Pressure Ductwork and Associated Components (grilles, etc.) associated with the outside air intake up to the CRB B2 None AQ-S

RG 1.78 N/A II radiation monitors downstream of the filter unit

RG 1.140

Charcoal and HEPA filtered

Maintain Positive Pressure Radiation Monitors (upstream of charcoal filter unit) CRB B2 None AQ

Backup diesel powered N/A III

Charcoal and HEPA filtered

Maintain Positive Pressure

CRV Filter Unit CRB B2 None AQ

RG 1.140 N/A III

CRV Supply Air Handling Unit A/B

Backup diesel powered

Ductwork and Associated Components (dampers, grilles, etc.) associated with the MCR or TSC

Charcoal and HEPA filtered

Isolation Dampers for CRV Filter Unit Bypass

Maintain Positive Pressure

CRV Battery Exhaust Fan A/B CRB B2 None AQ None N/A III

Temperature Sensors, Room Mounted RBVS, Reactor Building HVAC All components (except as listed below) RXB, RWB B2 None None None N/A III

RBV Supply AHUs RXB B2 None AQ

RG 1.140 N/A III

RBV General Area Exhaust Fans RWB

RBV General Area Exhaust Filter Units RWB

Hot Lab Exhaust Fan RXB Ductwork and Associated Components (Dampers, grilles, etc) (except for SFP exhaust components) RXB, RWB B2 None AQ None N/A III

RBV SFP Exhaust Ductwork and associated components (dampers, grills, etc.) RWB B2 None AQ

RG 1.140 N/A III

RBV SFP Exhaust Filter Units, including fans RWB Instrumentation RXB, RWB B2 None AQ

ANSI N13.1 N/A III

ANSI N42.18-2004

ANSIHPS N13.1-2001

Environmental Qualification

IEEE 497-2002 with CORR 1

RG 1.140

Table 1 of SRP 11.5 Tier 2 3.2-11 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

LRWS, Liquid Radioactive Waste System Degasifiers RXB B2 None AQ None RW-IIa RW-IIa LCW Collection Tanks RWB B2 None AQ None RW-IIb RW-IIb

Non-Radioactivity Indicating Instrumentation RWB, RXB B2 None None None N/A III

Drum Dryer RWB

LRW In-line Grab Samplers RWB, RXB Radioactivity Indicating Transmitter RWB, RXB B2 None AQ ANSI N42.18-2004 N/A III All other components RWB, RXB B2 None AQ None RW-IIc III GRWS, Gaseous Radioactive Waste System

Charcoal Guard Bed RWB B2 None AQ None RW-IIa RW-IIa

Charcoal Decay Beds

Charcoal Drying Heater RWB B2 None None None N/A III

Inlet Gas Sampler Radiation Indicating Transmitter RWB B2 None AQ ANSI N13.1-2011 N/A III All other components RWB B2 None AQ None RW-IIc III SRWS, Solid Radioactive Waste System Spent Resin Storage Tanks RWB B2 None AQ None RW-IIa RW-IIa Phase Separator Tanks RWB B2 None AQ None RW-IIb RW-IIb

Instrumentation RWB B2 None None None N/A III

Compactor

In-Line Grab Sampler All other components RWB B2 None AQ None RW-IIc III RWDS, Radioactive Waste Drain System All components RWB, RXB, B2 None None None D III ANB RWBVS, Rad-Waste Building HVAC System

Ductwork and Associated Components (Dampers, grilles, etc.) RWB B2 None AQ

RG 1.140 N/A III

RXB Exhaust Fan

Instrumentation

RWB Supply Air Handling Unit

RWB Supply Air Fans A/B All other components RWB B2 None None None N/A III MAE, Module Assembly Equipment

Module Inspection Rack RXB B2 None AQ-S None N/A II

Module Upender Module Import Trolley RXB B2 None None None N/A III MAEB, Module Assembly Equipment - Bolting RFT Support RXB B2 N/A Q None C I CNV Support Stand RXB B2 None AQ-S None N/A II All other components RXB B2 None None None N/A III FHE, Fuel Handling Equipment Fuel Handling Machine RXB B2 None AQ-S

ANSI/ANS 57.1-1992 N/A I

NUREG-0554

ASME NOG-1

New Fuel Elevator RXB B2 None AQ-S None N/A II

New Fuel Jib Crane Tier 2 3.2-12 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

SFSS, Spent Fuel Storage System Spent Fuel Storage Rack RXB B2 None AQ-S

ANSI/ANS 57.1-1992 N/A I

ANSI/ANS 57.2-1983 with additions, clarifications, and exceptions of RG 1.13

ANSI/ANS 57.3 SFPCS, Spent Fuel Pool Cooling System

Pumps RXB B2 None AQ ANSI/ANS 57.2-1983 with additions, D III

Strainers clarifications, and exceptions of RG 1.13

Valves - (PCUS boundary isolation valves)

Flow control orifices RXB B2 None None None N/A III

Instrumentation (pressure, temperature, flow, position)

All other components RXB B2 None None None D III PCUS, Pool Cleanup System All components (except as listed below) RXB B2 None AQ ANSI/ANS 57.2-1983 with additions, D III clarifications, and exceptions of RG 1.13 Instrumentation (Conductivity) RXB B2 None AQ ANSI/ANS 57.2-1983 with additions, N/A III clarifications, and exceptions of RG 1.13 Instrumentation (pressure, temperature, flow, position) RXB B2 None None None N/A III

Sample Points RXB B2 None None None D III

Instrumentation (pressure, temperature, flow, position)

RPCS, Reactor Pool Cooling System

Sample Points RXB B2 None AQ ANSI/ANS 57.2-1983 with additions, D III

Valves - (PCUS boundary isolation valves) clarifications, and exceptions of RG 1.13

Instrumentation - Boundary Valve Position RXB B2 None AQ ANSI/ANS 57.2-1983 with additions, N/A III clarifications, and exceptions of RG 1.13

Heat Exchangers RXB B2 None None None D III

Reactor Pool Cooling Pumps

Strainers

Valves (not listed above) - MOV, Air operated, Check, Manual, Relief

Instrumentation (not listed above) - Flow, Position, Pressure, Temperature RXB B2 None None None N/A III

Orifices Instrumentation - Temperature (PAM D Variable) RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I PSCS, Pool Surge Control System

RXB Penetrations - Piping RXB B2 None AQ-S None D II

Pool Penetrations - Piping RXB Tank Vent RE Yard B2 None AQ ANSI N42.18-2004 N/A III All other components RXB, Yard B2 None None None D III UHS, Ultimate Heat Sink UHS Pool (water only; also see RXB and RBCM below) RXB A1 N/A Q None N/A N/A Pool Level Instruments RXB B2 None AQ-S

IEEE 497-2002 with CORR 1 N/A I

NRC Order EA-12-051

NEI 12-02

NEI 12-06 (Order EA-12-049)

Water M/U Line RXB B2 None AQ-S

NRC Order EA-12-051 C I

NEI 12-02 PLDS, Pool Leakage Detection System All components RXB B2 None None None D III CES, Containment Evacuation System Vacuum Pump Suction Pressure Indicators RXB B2 None AQ-S None N/A I All other components (except as listed below) RXB B2 None AQ Quality Group D D III Tier 2 3.2-13 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

CES instrumentation (except as listed below) RXB B2 None None N/A N/A III Radiation Monitor RXB B2 None AQ

ANSI N42.18-2004 N/A III

ANSI/HPS N13.1-2011

Table 1 of SRP 11.5

Pressure boundary components of any monitoring path outside of containment shall be designed to withstand combustion events corresponding to the capability of containment.

Sample Vessel Radiation Transmitter RXB B2 None AQ

ANSI N42.18-2004 N/A III

Table 1 of SRP 11.5 Gas Discharge Radiation Transmitter RXB B2 None AQ

ANSI/HPS N13.1-2011 N/A III

Pressure boundary components of any monitoring path outside of containment shall be designed to withstand combustion events corresponding to the capability of containment.

PSS Sample Panel Inlet and Outlet Isolation Valves RXB B2 None AQ Pressure boundary components of any D III

Vacuum Pump Bypass Valve monitoring path outside of containment shall be designed to withstand combustion events corresponding to the capability of containment.

Charcoal Pre-Filter RXB B2 None AQ

RG 1.140 D III

Charcoal Filter

Discharge Filter

Containment Service Air Pressure Valve RXB B2 None None None D III

Sample Vessel Drain Sampler CFDS, Containment Flooding And Drain System All components (except as listed below) RXB B2 None None None D III CFD Module Post Accident Monitoring Return Valves RXB B2 None AQ Pressure boundary components of any D III monitoring path outside of containment shall be designed to withstand combustion events corresponding to the capability of containment.

Radiation Transmitter RXB B2 None AQ ANSI N42.18-2004 N/A III RCCWS, Reactor Component Cooling Water System All components (except as listed below) RXB B2 None None None D III Radioactivity Transmitters for: RXB B2 None AQ ANSI N42.18-2004 N/A III

RCCW CE Vacuum Pumps and Condensers

RCCW CVC NRHXs and PSS Coolers

RCCW PSS Cooling Water TCU RCCWS instrumentation RXB B2 None None None N/A III PSS, Process Sampling System All components (except as listed below) RXB, TGB B2 None None None N/A III Reactor coolant discharge sample line isolation valve RXB B2 None AQ ANSI N13.1 D III Primary sampling system analysis panel RXB B2 None AQ ANSI N13.1 N/A III Tier 2 3.2-14 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

Containment evacuation system sample line isolation valve RXB B2 None AQ

ANSI N13.1 D III

Pressure boundary components of any monitoring path outside of containment shall be designed to withstand combustion events corresponding to the capability of containment.

Containment sampling system sample panel RXB B2 None AQ

ANSI N13.1 D III

RG 1.7

Pressure boundary components of any monitoring path outside of containment shall be designed to withstand combustion events corresponding to the capability of containment.

Primary sampling system sample cooler cooling water chillers RXB B2 None AQ Quality Group D D III

Combined polisher effluents sample line isolation valve TGB B2 None None None D III

Condensate polisher sample line isolation valves

Condensate pump discharge sample line isolation valve

Condenser hotwell sample line isolation valve

Feedwater sample line isolation valves

Main Steam bypass sample line isolation valves

Main steam sample line isolation valves MSS, Main Steam System

Start-up Isolation Valves RXB B2 None AQ-S None D I

RXB Steam Traps

Secondary Main Steam Isolation Valves (Note 6) RXB B2 None AQ-S

Technical Specification Surveillance for D I

Secondary Main Steam Isolation Bypass Valves (Note 6) operability and inservice testing.

Valve Leak Detection

Secondary Main Steam Isolation Bypass Valve Close and Open Position Indicators RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

Secondary Main Steam Isolation Valve Close and Open Position Indicators

Auxiliary Steam Supply Valve TGB B2 None None None D III

Auxiliary Steam Warm-up Valve TGB

Main Steam Safety Valves Yard

Main Steam Vent Valve TGB

N2 Injection Isolation Valves RXB

Steam Sample Panel Isolation Valve TGB

TGB Steam Traps TGB

Main Steam Flow Transmitters RXB, TGB B2 None AQ

IEEE 497-2002 with CORR 1 N/A III

Main Steam Radiation Monitors

ANSI N42.18-2004 (Radiation Monitors)

Main Steam Pressure Transmitters RXB, TGB B2 None AQ None N/A III

Main Steam Temperature Elements All other components RXB, TGB B2 None None None N/A III FWS, Condensate and Feedwater System All components (except as listed below) TGB, RXB B2 None None None N/A III Feedwater Regulating Valve A/B (Note 6) RXB B2 None AQ-S Technical Specification Surveillance for D I operability and inservice testing.

Feedwater Supply Check Valve (Note 6) RXB B2 None AQ-S Inservice Testing D I Feedwater Regulating Valve Accumulators RXB B2 None AQ Technical Specification Surveillance for D III operability and inservice testing.

Feedwater Regulating Valve A/B Limit Switch RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

Condensate Storage Tank (located adjacent to TCB) Yard B2 None None None D III

Condensate Storage Tank Makeup Level Control Valve Tier 2 3.2-15 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

Steam Generator Differential Pressure Transmitter RXB B2 None AQ None N/A III

Condensate Header Emergency Rejection Level Control Valve TGB B2 None None None D III

Condensate Header Normal Rejection Level Control Valve

Condensate Polishing Rinse Recycle Pump Skid

Condensate Polishing System Inlet Thermal Well

Condensate Pump Liquid Seal Flow Orifice A/B/C

Condensate Pump Redundant Minimum Flow Protection valve

Condensate Pumps A/B/C

Condensate Filters A/B

Condensate Polishing Skid

Gland Steam Condenser Outlet Thermal Well

Condensate Strainers A/B

Feedwater Bypass Manual Valve

Feedwater Header Temperature Thermal Well

Feedwater Main Condenser

Feedwater Pumps A/B/C

Feedwater Pumps Minimum Flow Protection Control Valve A/B/C

Gland Steam Condenser Bypass Manual Valve

Long Cycle Cleanup AOV and Flow Control Valve

LP, IP, & HP Feedwater Heater

LP, IP, & HP FWH Inlet Thermal Well

LP, IP, & HP FWH Outlet Temperature Thermal Well

LP, IP, & HP FWH Outlet Thermal Well

LP/IP Feedwater Heater Bypass Manual Valve

Main Condenser Emergency Makeup Level Control Valve

Main Condenser Normal Makeup Level Control Valve

Main Condenser Thermal Well

PSS Sampler (Isolock)

Short Cycle Cleanup Flow Control Valve

Sparging Steam Control Valve FWTS, Feedwater Treatment All components (except as listed below) TGB B2 None None No D III CPRS, Condensate Polisher Resin Regeneration System All components TGB B2 None None

NEI 97-06 D III

EPRI PWR Secondary Water Chemistry Guidelines, Rev 7 HVDS, (Feedwater) Heater Vents and Drains System All components TGB B2 None None None D III CHWS, Chilled Water System All components RXB, CRB, B2 None None None N/A III CUB, RWB ABS, Auxiliary Boiler System

High Pressure and Low Pressure Aux Boiler skids ABB B2 None None No D III

High Pressure and Low Pressure Aux Boiler Condensate Tanks

High Pressure and Low Pressure Chemical Injection Packages

High Pressure Aux Boiler Flash Tank Radioactivity Instruments RXB, ABB B2 None AQ ANSI N42.18-2004 N/A III CARS, Condenser Air Removal System All components (except as listed below) TGB B2 None None None D III Tier 2 3.2-16 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

Effluent Radiation Element TGB B2 None AQ IEEE 497-2002 with CORR 1 N/A III

Effluent Radiation Transmitter

Discharge Flow Transmitter TGS, Turbine Generator System All components (except as listed below) TGB B2 None None None N/A III TG Gland Seal Exhauster Radiation Monitor TGB B2 None AQ ANSI N42.18-2004 N/A III TLOSS, Turbine Lube Oil Storage System All components TGB B2 None None None N/A III CPS, Cathodic Protection System Cathodic Protection System RXB, RWB, B2 None None None N/A III TGB, ANB, CRB, DGB, CUB, FWB, ABB, Yard, Other minor buildings CWS, Circulating Water System All components (except as listed below) TGB, Yard B2 None None None D III CWS pump bay and cooling tower basin level instrumentation TGB, Yard B2 None None None N/A III SCWS, Site Cooling Water System All components (except as listed below) RXB, CUB, B2 None None None D III TGB, ABB, Yard SCWS Instrumentation (except as listed below) RXB, CUB, B2 None None None N/A III TGB, ABB, Yard Letdown line rad monitor Yard B2 None AQ ANSI N42.18-2004 N/A III PWS, Potable Water System All components (except as listed below) Various B2 None None None N/A III Supply and return piping from the CRE penetration (includes only the isolation devices (loop seals) and CRB B2 None None None N/A II the piping between the loop seals and the outer wall of the CRE)

UWS, Utility Water System All components (except as listed below) Yard, RWB, B2 None None None N/A III FWB, RXB, TGB, CRB, ANB, CUB

Wastewater effluent discharge portion of UWS Yard B2 None None None D III

Discharge Basin

Local Grab Sample Line

Letdown Line Letdown Line Rad Monitor Yard B2 None AQ ANSI N42.18-2004 N/A III DWS, Demineralized Water System All components (except as listed below) Yard, RWB, B2 None None None D III RXB, ANB, CRB, ABB, TGB, CUB Radiation indication instruments for DWS headers RXB B2 None None None N/A III NDS, Nitrogen Distribution System All components Yard, RWB B2 None None None N/A III Tier 2 3.2-17 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

SAS, Service Air System All components CUB, ANB, B2 None None None N/A III RXB, TGB, RWB IAS, Instrument and Control Air System All components CUB, RWB, B2 None None None N/A III RXB, TGB, SCB, DGB, ANB TBVS, Turbine Building HVAC System All components TGB B2 None None None N/A III SBVS, Security Building HVAC System All components SCB B2 None None None N/A III DGBVS, Diesel Generator HVAC System All components DGB B2 None None None N/A III ABVS, Annex Building HVAC System All components ANB, RWB B2 None None None N/A III FPS, Fire Protection System All components CRB, RXB, TGB, B2 None AQ RG 1.189 N/A III RWB, SCB, ANB, DGB, ATB, FWB, WHB, CUB BPDS, BOP Drain System All components (except as listed below) TGB, CRB, B2 None AQ RG 1.26 D III CUB, DGB, ABB, FWB, Yard

Instrumentation TGB, CRB, B2 None None None N/A III

Radiation Monitor CUB, DGB, ABB, FWB, Yard EHVS, 13.8 KV and SWYD System All components TGB, Yard, B2 None None None N/A III Switchyard EMVS, Medium Voltage AC Electrical Distribution System All components TGB, RXB B2 None None None N/A III ELVS, Low Voltage AC Electrical Distribution System B6000 series Motor Control Centers RXB, CRB B2 None AQ None N/A III

Motor Control Center, non-B6000 RXB, CRB, TGB, B2 None None None N/A III

Station Service Transformers for B6000 and non-B6000 MCCs RWB, SCB,

Load Centers (SWG) for B6000 and non-B6000 MCCs ANB, ATB, CUB, Switchyard Tier 2 3.2-18 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

EDSS, Highly Reliable DC Power System

Channel A, Channel C, and Common Division I Components: RXB, CRB B2 None AQ-S

10 CFR 50.55a(1) N/A I

- DC Bus

10 CFR 50.55a(h)

- Switchgear

IEEE Std. 603-1991

- Batteries 1 and 2

Environmental Qualification

- Battery Chargers 1 and 2

Independence

- Transfer Switches 1 and 2

Single Failure Criterion

Channel B, Channel D, and Common Division II Components:

Common-Cause Failure

- DC Bus

Location of Indicators and Controls

- Switchgear

Multi-Unit Station Considerations

- Batteries 1 and 2

- Battery Chargers 1 and 2

- Transfer Switches 1 and 2

EDSS-C, Cabling

EDSS-C, Fusible Disconnects

EDSS-MS, Cabling

EDSS-MS, Fusible Disconnects

Channel A, Channel C, and Common Division I Components: RXB, CRB B2 None AQ-S

10 CFR 50.55a(1) N/A I

- Battery Charger Ammeters 1 and 2

10 CFR 50.55a(h)

- Battery Monitors 1 and 2

IEEE Std. 603-1991

- DC Bus Ground Fault Relay

Environmental Qualification

- DC Bus Overvoltage Relay

Independence

- DC Bus Undervoltage Relay

Single Failure Criterion

Channel B, Channel D, and Common Division II Components:

Common-Cause Failure

- Battery Charger Ammeters 1 and 2

Location of Indicators and Controls

- Battery Monitors 1 and 2

Multi-Unit Station Considerations

- DC Bus Ground Fault Relay

- DC Bus Overvoltage Relay

- DC Bus Undervoltage Relay Channel A, Channel B, Channel C, Channel D, Common Division I, and Common Division II DC Bus RXB, CRB B2 None AQ-S

10 CFR 50.55a(1) N/A I Voltmeters

10 CFR 50.55a(h)

IEEE Std. 603-1991

Environmental Qualification

Independence

Single Failure Criterion

Common-Cause Failure

Location of Indicators and Controls

Multi-Unit Station Considerations

IEEE 497-2002 with CORR 1 EDNS, Normal DC Power System All components RXB, CRB, B2 None None None N/A III RWB, TGB, Yard BPSS, Backup Power Supply System All components (except as listed below) DGB, RXB, B2 None AQ-S None N/A II Yard Auxiliary AC Power Supply Yard B2 None None None N/A III PLS, Plant Lighting System All components (except as listed below) All Buildings B2 None None None N/A III Main Control Room DC emergency lighting (including fixtures, cables, and lighting boards) CRB B2 None AQ

Powered from highly-reliable DC power N/A III distribution system

Environmental Qualification Tier 2 3.2-19 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

GLPS, Grounding and Lightning Protection System All components RXB, TGB, B2 None None None N/A III RWB, SCB, ANB, DGB, ATB, CUB, FWB, CRB SPS, Security Power System All components Various B2 None None None N/A III MPS, Module Protection System All components (except as listed below) RXB, CRB A1 N/A Q None N/A I

Division I and Division II Engineered Safety Features Actuation System: RXB, CRB A2 N/A Q None N/A I

- Equipment Interface Modules for Secondary MSIVs, Secondary MSIV Bypass Isolation Valves and Feedwater Regulating Valves for Containment Isolation and DHRS Actuation

Manual LTOP Actuation Switch

Separation Group A, B, C, and D:

- Safety Function Module and associated Maintenance Switch for LTOP function

Separation Group A - Safety Function Module: RXB B2 None AQ-S

IEEE 497-2002 with CORR 1 N/A I

- Feedwater Indication and Control

EMI/RFI

- Leak Detection into Containment

Environmental Qualification

Separation Group B and C - Safety Function Module for PAM indication functions

Power from Vital Instrument Bus

Separation Group D - Safety Function Module:

10 CFR 50.55a(1)

- Leak Detection into Containment

10 CFR 50.55a(h)

IEEE Std. 603-1991

Independence

Single Failure Criterion

Common-Cause Failure

Location of Indicators and Controls

Multi-Unit Station Considerations

24-Hour Timers for PAM-only Mode RXB B2 None AQ-S

IEEE 497-2002 with CORR 1 N/A I

Division I and Division II:

EMI/RFI

- Engineered Safety Features Actuation System - Equipment Interface Module for low AC voltage to

Environmental Qualification battery chargers function

Power from Vital Instrument Bus

- Engineered Safety Features Actuation System Monitoring and Indication Bus, Communication Module

- MPS Gateway

- Reactor Trip System Monitoring and Indication Bus - Communication Module

Separation Group A, B, C, and D:

- Monitoring and Indication Bus - Communication Module

Separation Group B and C - Safety Function Modules for PAM indication functions Division I and II Maintenance Workstations RXB B2 None AQ-S None N/A II NMS, Neutron Monitoring System

Excore Neutron Detectors RXB A1 N/A Q None N/A I

Excore Separation Group A/B/C/D - Power Isolation, Conversion and Monitoring Devices

Excore Signal conditioning and processing equipment

Flood Highly Sensitive Neutron Detectors (for CNV flooding events) RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

Flood Signal conditioning and processing equipment (for CNV flooding events)

Refuel Neutron Detectors (for refueling) RXB B2 None AQ-S None N/A II

Refuel Signal conditioning and processing equipment (for refueling)

SDIS, Safety Display and Indication System All components CRB B2 None AQ-S

IEEE 497-2002 with CORR 1 N/A I

EMI/RFI

Power from Vital Instrument Bus Tier 2 3.2-20 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

MCS, Module Control System

RSS HMI RXB, CRB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A II

MCR HMI

MCS Domain Controller (Green)

MCS Domain Controller (Yellow)

Gateway from MPS RXB, CRB B2 None AQ-S None N/A II

Gateway to PCS

Cabinets (PAM E Variables) RXB, CRB, TGB B2 None AQ IEEE 497-2002 with CORR 1 N/A III

Controllers (PAM E Variables)

I/O Modules (PAM E Variables)

Controllers (other than above) RXB, CRB, TGB B2 None AQ None N/A III

I/O Modules (other than above)

ICIS, In-Core Instrumentation System In-core instrument string sheath RXB A2 N/A Q None A I In-core instrument string/ temperature sensors RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I In-core instrument string/ flux sensors RXB B2 None AQ-S None N/A I Signal Conditioning and Processing Electronics RXB B2 None AQ-S None N/A II PCS, Plant Control System

Controllers CRB, RXB, TGB, B2 None AQ

Backup diesel powered N/A III

I/O Modules RWB

Analyzed for seismic qualification

Controllers for RSS indication RXB B2 None AQ IEEE 497-2002 with CORR 1 N/A III

I/O Modules for RSS indication

Cabinets CRB, RXB, TGB, B2 None AQ

IEEE 497-2002 with CORR 1 N/A III

PCS Domain Controller (Green) RWB

Backup diesel powered

PCS Domain Controller (Yellow)

Analyzed for seismic qualification

RSS HMI

MCR HMI

Gateway from MCS X CRB, RXB, B2 None None None N/A III

Gateway from PPS RWB

RWBCR HMI PPS, Plant Protection System

Division I and Division II: CRB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I

- Monitoring and Indication Bus Communication Modules

- Division I Safety Function Module for Spent Fuel Pool and Reactor Pool Level Indication

- Equipment Interface Modules:

CRH Air Supply Isolation Valve

CRH Pressure Relief Isolation Valve

CRV General Exhaust Damper

CRV Return Air Damper

CRV Smoke Purge Exhaust Damper

CRV Supply Air Damper

Division I and Division II Safety Function Module for EDSS-C Bus Voltage Indication Division I and Division II: CRB B2 None AQ-S None N/A I

ELVS Voltage Sensors

Manual CRH Actuation Switches Division I and Division II Safety Function Module for CRE Air Flow Delivery Indication CRB B2 None AQ-S None N/A I Tier 2 3.2-21 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

Division I and Division II: CRB B2 None AQ-S RG 1.78 N/A I

CTB Communication Module

Enable Nonsafety Control Switch

Hard-Wired Module

Scheduling and Bypass Modules

Safety Function Modules for CRV Post-filter Radiation Sensor

Safety Function Module for CRV Post-filter Radiation Sensor Trip/Bypass Switches Division I and Division II: CRB B2 None AQ-S RG 1.78 N/A I

CRV Outside Air Isolation Damper Equipment Interface Module

Manual Outside Air Isolation Actuation Switch

Safety Function Module for CRV Toxic Gas Sensor

Safety Function Module for CRV Toxic Gas Sensor Trip/Bypass Switch Division I and Division II Maintenance Workstations CRB B2 None AQ-S None N/A II RMS, Radiation Monitoring System RM system that monitors PAM B & C variables RXB B2 None AQ-S IEEE 497-2002 with CORR 1 N/A I Radiation monitors that monitors Type E variables RXB, TGB B2 None AQ IEEE 497-2002 with CORR 1 N/A III Area airborne radiation monitors that monitors Type E Variable CRB, RXB B2 None AQ

IEEE 497-2002 with CORR 1 N/A III

ANSI/HPS N13.1-2011 Area airborne radiation monitors in: ANB, RWB, B2 None AQ ANSI/HPS N13.1-2011 N/A III

Annex Building RXB

Radioactive Waste Building

Reactor Building Radiation monitors in: ANB, CRB, B2 None AQ None N/A III

Annex Building RWB, RXB,

Control Building TGB

Radioactive Waste Building

Reactor Building

Turbine Buildings RXB, Reactor Building Reactor Building (includes interior walls and floor forming UHS pool) Yard A1 N/A Q None N/A I RBC, Reactor Building Cranes Reactor Building Crane RXB B1 None AQ-S ASME NOG-1 N/A I Module Lifting Adapter RXB B1 None AQ-S ANSI N14.6 N/A N/A Traveling Jib Crane RXB B2 None N/A None N/A II Wet Hoist RXB B2 None AQ ASME NOG-1 N/A N/A RBCM, Reactor Building Components Over-Pressurization Vents (OPV) RXB A2 None Q None C(d) I

UHS Pool Liner and Dry Dock Liner RXB B2 None AQ-S ANSI/ANS 57.2-1983 with additions, N/A I Dry Dock Gate support stainless steel plates at plate-to-liner weld locations clarifications, and exceptions of RG 1.13 Bioshield RXB B2 None AQ-S EQ requirements to GDC 4 and 23 N/A II Reactor Building Equipment Door RXB B2 None AQ-S None N/A II Dry Dock Gate RXB B2 None AQ-S None N/A II

Dry Dock Gate Closure instrumentation RXB B2 None None None N/A III

Reactor Building Equipment Door Condition Instrumentation

TGB, Turbine Generator Building

Turbine Generator Building Yard B2 None None None N/A III

TBC, Turbine Building Cranes

Turbine Building Cranes TGB B2 None None None N/A III RWB, Radioactive Waste Building Radioactive Waste Building Yard B2 None AQ None RW-IIa II, RW-IIa Tier 2 3.2-22 Revision 4

NuScale Final Safety Analysis Report Classification of Structures, Systems, and Components Table 3.2-1: Classification of Structures, Systems, and Components (Continued)

SSC (Note 1) Location SSC Classification RTNSS Category QA Program Augmented Design Requirements Quality Group / Safety Seismic Classification (A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification (Ref. RG 1.29 or RG 1.143)

(Note 2) (Ref RG 1.26 or RG 1.143) (Note 5)

(Note 4)

SCB, Security Buildings (Guardhouse)

Security Building Yard B2 None None None N/A III

Vehicle inspection sally port

ANB, Annex Building

Annex Building Yard B2 None None None N/A III

DGB, Diesel Generator Building

Diesel Generator Building Yard B2 None None None N/A III

CUB, Central Utility Building

Central Utility Building Yard B2 None None None N/A III

FWB, Firewater Building

Firewater Building Yard B2 None None None N/A III CRB, Control Building CRB Structure at EL 120-0 and below (except as discussed below). Yard A1 N/A Q None N/A I

CRB Structure above EL 120-0 Yard B2 None AQ-S None N/A II

Inside the CRB elevator shaft and two stairwells, full height of structure

CRB Fire Protection Vestibule (on East Side of CRB)

MEMS, Metrology and Environmental Monitoring System All components Yard, CRB B2 None AQ IEEE 497-2002 with CORR 1 N/A III COMS, Communication Systems All components Yard for B2 None None None N/A III collection of data CRB for display of results SMS, Seismic Monitoring System All components RXB, CRB B2 None AQ-S None N/A I Note 1: Acronyms used in this table are listed in Table 1.1-1.

Note 2: QA Program applicability codes are as follows:

Q = indicates quality assurance requirements of 10 CFR 50 Appendix B are applicable in accordance with the quality assurance program (see Section 17.5).

AQ = indicates that pertinent augmented quality assurance requirements for nonsafety-related SSCs are applied to ensure that the function is accomplished when needed based on that functionality's regulatory requirements. Note that in meeting regulatory guidance, codes, and standards, those applicable SSCs may also have quality assurance requirements invoked by said guidance (e.g., RG 1.26, RG 1.143, IEEE 497, RG 1.189).

AQ-S = indicates that the pertinent requirements of 10 CFR 50 Appendix B are applicable to nonsafety-related SSC classified as Seismic Category I or Seismic Category II in accordance with the quality assurance program.

None = indicates no specific QA program or augmented quality requirements are applicable.

Note 3: Additional augmented design requirements, such as the application of a Quality Group, radwaste safety, or seismic classification, to nonsafety-related SSC are reflected in the columns Quality Group/Safety Classification and Seismic Classification, where applicable.

Note 4: See Section 3.2.2.1 through Section 3.2.2.4 for the applicable codes and standards for each RG 1.26 Quality Group designation A, B, C, and D. A Quality Group classification per RG 1.26 is not applicable to supports or instrumentation. See Section 3.2.1.4 for a description of RG 1.143 classifications for RW-IIa, RW-IIb, and RW-IIc.

Note 5: Where SSC (or portions thereof) as determined in the as-built plant which are identified as Seismic Category III in this table could, as the result of a seismic event, adversely affect Seismic Category I SSC or result in incapacitating injury to occupants of the control room, they are categorized as Seismic Category II consistent with Section 3.2.1.2 and analyzed as described in Section 3.7.3.8.

Note 6: Provides nonsafety-related backup isolation to a safety-related isolation device. See FSAR Sections 3.9.6.5, 15.0.0.6.6 and Table 3.9-17.

Note 7: Includes all subcomponents of the reactor vessel internals upper riser assembly with the exception of the bellows lateral seismic restraining structure and bellows vertical expansion structure which are listed separately.

Tier 2 3.2-23 Revision 4

The design includes three structures that are evaluated for wind and tornado loadings: the Seismic Category I Reactor Building (RXB) and Control Building (CRB) [the CRB is Seismic Category II above elevation 120' and in the areas below 120' defined in Section 1.2.2.2] and the Seismic Category II Radioactive Waste Building (RWB). The RXB, CRB and RWB are enclosed structures. This section describes the design approach for severe and extreme wind loads on these structures. Section 3.8.4 discusses the design of the Seismic Category I Structures.

The Seismic Category II RWB is also classified as RW-IIa (High Hazard) in accordance with Regulatory Guide (RG) 1.143, Rev. 2, "Design Guidance For Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants."

The RWB is designed using the same wind, tornado and hurricane loads as specified for as the Seismic Category I structures. This meets or exceeds the wind load specified in Table 2 of RG 1.143, Rev. 2. This regulatory guide directs the use of ASCE 7-95 for wind loads. However, ASCE 7-05 (Reference 3.3-1) is used for wind loads in this design. Similarly, the tornado missiles from RG 1.76, Rev.1, "Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants," are used rather than the tornado missiles identified in Table 2 of RG 1.143, Rev. 2.

In addition, other structures, systems, and components that have the potential to interact with the Seismic Category I buildings are evaluated to demonstrate they do not adversely affect the RXB or Seismic Category I portions of the CRB. This is described in Section 3.3.3.

The design complies with General Design Criteria 2 and 4 in that structures, systems, and components are designed to withstand the most severe effects of natural phenomena wind, hurricane, and tornadoes without loss of the capability to perform their safety functions. This is achieved by establishing design parameters that are representative of a reasonable number of potential plant site locations in the United States. Design parameters for severe wind loads are provided in Section 3.3.1.1 and design parameters for extreme wind loads are provided in Section 3.3.2.1.

The RWB has been evaluated for severe and extreme wind loads using the methodology in Section 3.3.1.2 and Section 3.3.2.2 and can withstand the severe and extreme winds.

1 Severe Wind Loadings 1.1 Design Parameters for Severe Wind The design basis severe wind is a 3-second gust at 33 feet above ground for exposure category C. The wind speed (Vw) is 145 mph. The wind speed is increased by an importance factor of 1.15 for the design of the RXB, CRB, and RWB. These design parameters are based upon ASCE/SEI 7-05.

1.2 Determination of Severe Wind Forces The maximum velocity pressure (qz) based on the applicable maximum wind speed (Vw) is calculated in conformance with ASCE/SEI 7-05 (Reference 3.3-1), Equation 6-15, as follows:

2 3.3-1 Revision 4

where, Kz = velocity pressure exposure coefficient evaluated at height "z", as defined in ASCE/SEI 7-05, Table 6-3, but not less than 0.87. For simplicity and conservatism, z is assumed to be the building height, Kzt = topographic factor equal to 1.0, Kd = wind directionality factor equal to 1.0, Vw = maximum wind speed equal to 145 mph, and I = importance factor equal to 1.15 for the RXB, CRB, and RWB.

Design wind loads on the RXB, CRB, and RWB are determined in conformance with ASCE/SEI 7-05 (Reference 3.3-1), Equation 6-17:

p=qGCp - qi (GCpi) (lb/ft2) where, G = gust factor equal to 0.85, Cp = external pressure coefficient equal to 1.0, GCpi = internal pressure coefficient equal to 0.18, q = velocity pressure, and qi = internal velocity pressure.

2 Extreme Wind Loads (Tornado and Hurricane Loads) 2.1 Design Parameters for Extreme Winds Tornado wind loads include loads caused by the tornado wind pressure, tornado atmospheric pressure change effect, and tornado-generated missile impact. Hurricane wind loads include loads due to the hurricane wind pressure and hurricane-generated missiles.

The parameters for the design basis tornado are the most severe tornado parameters postulated for the contiguous United States as identified in RG 1.76, Rev. 1.

Maximum wind speed . . . . . . . . . . . . . . . . . . . . 230 mph

Translational speed . . . . . . . . . . . . . . . . . . . . . . . 46 mph 2 3.3-2 Revision 4

Radius of maximum rotational speed . . . . . . 150 ft

Pressure drop. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2 psi

Rate of pressure drop . . . . . . . . . . . . . . . . . . . . . 0.5 psi/s The wind speed for the design basis hurricane is the highest wind speed postulated in Regulatory Position 1 of RG 1.221, Rev. 0, "Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants," which occurs in Figure 2 of RG 1.221, Rev. 0.

Maximum wind speed . . . . . . . . . . . . . . . . . . . . 290 mph Refer to Section 3.5 for a description of hurricane and tornado wind-generated missiles.

2.2 Determination of Tornado and Hurricane Forces Tornado and hurricane wind velocities are converted into effective pressure loads in accordance with ASCE/SEI 7-05 (Reference 3.3-1), Equation 6-15, as follows:

qz=0.00256 Kz Kzt Kd Vw2 I (lb/ft2) where, Kz = velocity pressure exposure coefficient evaluated at height "z", as defined in with ASCE/SEI 7-05, Table 6-3, but not less than 0.87. (For tornados, wind speed is not assumed to vary with height.) For simplicity and conservatism, z is assumed to be the building height.

Kzt = topographic factor equal to 1.0, Kd = wind directionality factor equal to 1.0, Vw = maximum wind speed (mph) (For tornadoes, Vw is the resultant of the maximum rotational speed and the translational speed), and I = importance factor equal to 1.15 for the RXB, CRB, and RWB.

Extreme wind loads on the RXB, CRB, and RWB are determined in conformance with ASCE/SEI 7-05, Equation 6-17:

p=qGCp - qi (GCpi) (lb/ft2) where, G = gust factor equal to 0.85, Cp = external pressure coefficient equal to 1.0, 2 3.3-3 Revision 4

q = velocity pressure, and qi = internal velocity pressure.

Internal pressure from the tornado is the design parameter for maximum pressure drop.

2.3 Combination of Forces The most adverse of the following combinations are considered for the total hurricane or tornado load:

Wt = Wp Wt = Ww + 0.5 Wp + Wm where, Wt = total load, Ww = load from wind effect, Wp = load from tornado atmospheric pressure change effect (Wp = 0 for hurricanes),

and Wm = load from missile impact effect.

Item 3.3-1: A COL applicant that references the NuScale Power Plant design will confirm that nearby structures exposed to severe and extreme (tornado and hurricane) wind loads will not collapse and adversely affect the Reactor Building or Seismic Category I portion of the Control Building.

3 References 3.3-1 American Society of Civil Engineers/Structural Engineering Institute, "Minimum Design Loads for Buildings and Other Structures," ASCE/SEI 7-05, Reston, VA.

2 3.3-4 Revision 4

Flooding of a nuclear power plant can come from internal sources - piping ruptures, tank failures or the actuation of fire suppression systems, or from external sources - flooding from nearby water bodies or precipitation. Section 3.4.1 evaluates flooding effects of discharged fluid resulting from the high and moderate energy line breaks and cracks; from fire-fighting activities; and from postulated failures of non-seismic and non-tornado protected piping, tanks, and vessels outside the structures. In the absence of final pipe routing information, the flooding hazards are representative of the flooding hazards expected throughout the plant.

The design satisfies General Design Criterion 4 in that the structures, systems, and components (SSC) are designed to withstand the effects of environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents without loss of the capability to perform their safety functions.

The design also satisfies General Design Criterion 2 in that SSC accommodate the effects of natural phenomena, including floods, without losing the ability to perform their safety function. Section 3.4.2 addresses flooding from natural phenomena.

Dynamic effects from pipe rupture are addressed in Section 3.6. Environmental effects are addressed in Section 3.11. Loads on Seismic Category I and other structures are addressed in Section 3.8.

1 Internal Flood Protection for Onsite Equipment Failures Internal flooding analyses were performed in the Reactor Building (RXB) and the Control Building (CRB) to confirm that flooding from postulated failures of tanks and piping or actuation of fire suppression systems does not cause the loss of equipment required to: (a) maintain the integrity of the reactor coolant pressure boundary for any module, (b) shut down the reactor for any module and maintain it in a safe shutdown condition, or (c) prevent or mitigate the consequences of accidents which could result in unacceptable offsite radiological consequences. These SSC are collectively identified as "equipment subject to flood protection."

Table 3.4-2 identifies the rooms that contain SSC that have safety-related or risk-significant attributes that are subject to flood protection. The flooding analysis considers areas and rooms that contain these SSC, not the specific SSC themselves. Safety-related cable is either routed above the flood level or qualified for submergence. Rooms where cable is the only safety-related SSC are not included. Mitigation of flooding in the identified rooms will be accomplished by, for example, watertight or water resistant doors, elevating equipment above the flood level, enclosing or qualifying equipment for submersion, or other similar type of flood protection.

The internal flooding analysis is conducted on a level-by-level and room-by-room basis for the Seismic Category I RXB and CRB for the postulated flooding events.

The RXB and CRB flooding analysis consists of the following steps:

identification of potential flooding sources

identification of rooms/areas that contain equipment subject to flood protection 2 3.4-1 Revision 4

determination of the need for protection and mitigation measures for rooms containing equipment subject to flood protection 1.1 Assumptions used in the Flooding Analyses Unless a stress analysis has been performed to identify potential break locations or eliminate the piping from consideration of potential breaks, high and moderate energy piping greater than 2 inches nominal diameter are assumed to have a full circumferential break in any room or area where they pass. The design operational pressure/flow rate is used to estimate leakage flow rates. The total quantity of fluid released is consistent with the action necessary to isolate the line. The following assumptions are used for isolation times:

For the CRB:

Thirty minutes are assumed between leak initiation and leak isolation (the CRB is continuously occupied).

For the RXB:

Thirty minutes are assumed between initiation of a leak and detection by any means (except for the main steam line which automatically isolates).

Ten minutes are assumed between leak detection and isolation.

Fire suppression activities are also a potential flooding source. The discharge of the fire suppression system for the RXB and CRB is assumed to be 700 gpm and 550 gpm, respectively. These estimates are based on the automatic fire suppression flow rate of 0.3 gpm/ft2 over a 1,500 ft2 area for the RXB and 0.2 gpm/ft2 over a 1,500 ft2 area for the CRB based on the occupancy categories of NFPA 13 (Reference 3.4-1) with the addition of 250 gpm for manual hose flow (NFPA 14, Reference 3.4-2). The fire suppression duration is assumed to be two hours for the RXB and 60 minutes for the CRB based on the occupancy categories of NFPA 13.

The following assumptions are used to determine flood water volumes in rooms and areas within the RXB and CRB:

Floor drains and sump pumps are not credited for reducing flood water volume during the event.

Backflow through floor drains is not considered. It is assumed to be bounded by the direct flooding pathways. Floor drains are discussed in Section 9.3.3.

Interior doors, unless specified as a watertight/waterproof door, are assumed to fail open or provide a high leak flow rate between rooms.

In areas with multiple sources, each source is considered separately.

1.2 Reactor Building Flooding Analysis There are multiple flooding sources in the RXB. The sources are discussed below, and the water sources and volumes are listed in Table 3.4-1.

2 3.4-2 Revision 4

fire protection water to the fire suppression sprinkler system on each RXB elevation. A break in the fire protection line can provide up to 2500 gpm from the pipe rupture. The water from the rupture is assumed to be released for 40 minutes.

Fire suppression activities consisting of area sprinklers and operating fire hoses with a flowrate of 700 gpm total (450 gpm + 250 gpm respectively), are assumed to provide flooding water for two hours.

Reactor Building HVAC system chilled water cooling coil piping (from Site Cooling Water) has a flow of 1,000 gpm that is assumed to provide flood water for 40 minutes.

The site cooling water system header piping into the RXB at elevation 100' has a flow of 5,000 gpm that is assumed to provide flood water for 40 minutes.

Demineralized water system and utility water system has a flowrate of 300 gpm.

The pipe rupture is assumed to provide floodwater for 40 minutes.

Main steam line break has such a small time frame between the break and pipe isolation (five seconds) that the condensed steam from the break will not cause an internal flood.

Feedwater line break has a flow of 600 gpm that is assumed to provide flood water for 40 minutes.

The spent fuel pool cooling and reactor pool cooling inlet and outlet piping are routed from elevation 85' to elevation 50'. A break in either piping line on elevation 75' or elevation 50' could drain 158,900 ft3 and result in a flood height of 9'-4 3/4" on elevation 50' and 14'-7 1/2" on elevation 75'. Each of the rooms that contain SSC subject to flood prevention either have flood doors or the equipment in the rooms are designed or protected for submergence.

Chemical volume control system (CVCS) line break has a flow of 90 gpm that is assumed to provide flood water for 40 minutes.

Pool surge control system line break has a maximum flow of 2747 gpm that is assumed to provide flood water for 40 minutes.

Auxiliary boiler system has maximum break flow of 80 gpm and that is assumed to provide flood water for 40 minutes.

1.2.1 Flooding at Elevation 125'-0" Flooding of this elevation results from a fire suppression system actuation or a site cooling water pipe break. The electrical and mechanical equipment rooms on this elevation contain SSC that are subject to flood protection. Water level on this elevation is predicted to be less than four inches. Individual rooms subject to flood protection are shown in Table 3.4-2.

1.2.2 Flooding at Elevation 100'-0" Flooding of this elevation can be caused by fire suppression system actuation or a feedwater line break. The feedwater line break produces the highest water level of 2 3.4-3 Revision 4

1.2.3 Flooding at Elevation 86'-0" A fire suppression system actuation in the hallways provides flooding water for this elevation. However, the metal floor grating in the hallways allows the flood water to drain to elevation 75'-0".

1.2.4 Flooding at Elevation 75'-0" Elevation 75'-0" of the RXB contains the remote shutdown station and other electrical equipment rooms that house SSC that are subject to flood protection.

Grating in elevation 86'-0" hallway floors drains flood water from that elevation to elevation 75'-0" hallways. However, fire suppression activities in the elevation 75-0 hallways produces the highest flood level of approximately 23 inches.

Individual rooms containing equipment subject to flood protection have smaller flood levels. Individual rooms subject to flood protection are shown Table 3.4-2.

1.2.5 Flooding at Elevation 62'-0" Miscellaneous mechanical equipment rooms are located on elevation 62-0. There are no SSC subject to flood protection located at this elevation.

1.2.6 Flooding at Elevation 50'-0" Elevation 50'-0" contains CVCS equipment, demineralized water valves, and miscellaneous mechanical and electrical equipment rooms. Fire suppression activities in the hallways produces the highest flood level of approximately 16.5 inches.

1.2.7 Flooding at Elevation 35'-8" Elevation 35'-8" contains CVCS pump rooms and miscellaneous mechanical equipment rooms. There are no SSC subject to flood protection located at this elevation.

1.2.8 Flooding at Elevation 24'-0" Elevation 24'-0" contains CVCS filters and ion exchangers and miscellaneous mechanical equipment rooms. There are no SSC subject to flood protection located on this elevation.

1.2.9 Containment Flooding Analysis Containment is flooded as part of normal shutdown, and may also be flooded as part of accident mitigation as described in Chapter 15. Therefore, there is no equipment subject to flood protection inside containment and no containment flooding analysis is necessary.

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There are four potential flooding sources in the CRB. The sources are discussed below, and the water volumes and sources are listed in Table 3.4-1.

The 6-inch fire protection main line enters the CRB through the fire riser room between the 100' and 120' floor level. From this header, the pipe distributes fire protection water to the fire suppression sprinkler system located on each CRB elevation. A break in the fire protection line can provide up to 2,225 gpm from the pipe rupture. The water from the rupture is assumed to be released for 30 minutes.

The 4-inch chilled water supply provides water to the HVAC system on elevation 120' of the CRB, and has a flow of 226 gpm that is assumed to provide flood water for 30 minutes.

The 2-inch potable water supply pipe provides potable water to floor elevation 76' 6" and elevation 100'. Though this line is not considered a large pipe, its routing through the CRB poses a flooding risk. The system has a flow of 50 gpm that is assumed to provide flood water for 30 minutes.

Fire suppression activities consisting of area sprinkler and operation fire hoses with a flow rate of 550 gpm (300 gpm + 250 gpm, respectively), are assumed to provide flooding water for one hour.

1.3.1 Flooding at Elevation 120'-0" Elevation 120'-0" contains HVAC and miscellaneous mechanical equipment. There are no SSC subject to flood protection located at this elevation.

1.3.2 Flooding at Elevation 100'-0" Flooding at the 100'-0" elevation could occur from a break in the potable water system, a break in the fire suppression riser, or from fire-fighting activities. There are no SSC subject to flood protection at elevation 100'-0".

The fire riser room is located outside the main building next to the vestibule. The fire riser is a potential flooding source in the CRB. However, the water from the riser will flow into the vestibule and out to the environment or into the main hallway and down the stairwells and will have no impact on elevation 100-0.

1.3.3 Flooding at Elevation 76'-6" The main control room is located on elevation 76'-6". This room contains equipment subject to flood protection. Flooding could occur from actuation of the sprinkler system in an adjacent hallway or from a break in the potable water line that is routed which is in rooms connected to the hallway. Due to the small volume of water from a potable water system line break, sprinkler actuation is the dominant flooding source. Firefighting activities in the adjacent rooms could result in a flood depth of approximately 17.5 inches.

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Elevation 63'-3 contains electrical equipment and utility rooms. There are no SSC subject to flood protection located at this elevation.

1.3.5 Flooding at Elevation 50'-0" Elevation 50'-0" contains electrical equipment, air bottles, and utility rooms. There are no SSC that are subject to flood protection at this elevation.

1.4 Flooding Outside the Reactor and Control Buildings Flooding of the RXB or CRB caused by external sources does not occur. The design external flood level is established as less than 99' elevation (one foot below the baseline plant elevation (top of concrete) at 100'-0"). The finished grade at the building perimeter of the RXB and CRB is approximately 6 inches below the top of concrete elevation, except at the CRB tunnel and a truck ramp on the south side of the Radwaste Building.

Water from tanks and piping that are non-seismic and non-tornado/hurricane protected is a potential flooding source outside the buildings. However, there are no large tanks or water sources near the entrances to the CRB and RXB. The site is graded to transport water away from these buildings. Therefore, failure of equipment outside the CRB and RXB cannot cause internal flooding.

1.5 Site Specific Analysis Item 3.4-1: A COL applicant that references the NuScale Power plant design certification will confirm the final location of structures, systems, and components subject to flood protection and final routing of piping.

Item 3.4-2: A COL applicant that references the NuScale Power plant design certification will develop the on-site program addressing the key points of flood mitigation. The key points to this program include the procedures for mitigating internal flooding events; the equipment list of structures, systems, and components subject to flood protection in each plant area; and providing assurance that the program reliably mitigates flooding to the identified structures, systems, and components.

Item 3.4-3: A COL applicant that references the NuScale Power plant design certification will develop an inspection and maintenance program to ensure that each water-tight door, penetration seal, or other degradable measure remains capable of performing its intended function.

Item 3.4-4: A COL applicant that references the NuScale Power plant design certification will confirm that site-specific tanks or water sources are placed in locations where they cannot cause flooding in the Reactor Building or Control Building.

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The design includes the two Seismic Category I structures: the RXB and the CRB. The Radioactive Waste Building (RWB) is Seismic Category II and does not contain any equipment subject to flood protection. There are no other safety-related structures in the design.

2.1 Probable Maximum Flood The design is the equivalent of a "Dry Site" as defined in Regulatory Guide 1.102, "Flood Protection for Nuclear Power Plants," Rev. 1. The Seismic Category I structures are protected from external floods and groundwater by establishing the following design parameters:

The probable maximum flood elevation (including wave action) of the design is one foot below the baseline plant elevation (100-0).

The maximum groundwater elevation for the design is two feet below the baseline plant elevation.

With the exceptions of the subgrade CRB tunnel and a truck ramp on the south side of the Radwaste Building, the finished grade for all building structures is approximately six inches below the baseline plant elevation. The yard is graded with a minimum slope of 1.5 percent away from these structures.

The below grade portions of the Seismic Category I structures provide protection for the safety-related and risk-significant SSC from groundwater intrusion by utilizing the following design features:

the portions of the buildings that are below grade consider the use of waterstops and waterproofing

exterior below grade wall or floor penetrations have watertight seals

waterproofing and dampproofing systems, if used, are applied per the International Building Code Section 1805 (Reference 3.4-3)

waterproofing and dampproofing materials, if used in horizontal applications, will have a coefficient of static friction equal to or greater than the site parameter established in Table 2.0-1 for all interfaces between the basemat and soil.

The design does not use a permanent dewatering system.

Item 3.4-5: A COL applicant that references the NuScale Power Plant design certification will determine the extent of waterproofing and dampproofing needed for the underground portion of the Reactor Building and Control Building based on site-specific conditions. Additionally, a COL applicant will provide the specified design life for waterstops, waterproofing, damp proofing, and watertight seals. If the design life is less than the operating life of the plant, the COL applicant will describe how continued protection will be ensured.

Item 3.4-7: A COL applicant that references the NuScale Power Plant design certification will determine the extent of waterproofing and damp proofing needed to prevent 2 3.4-7 Revision 4

corresponding Reactor Building connecting walls.

The NuScale Power Plant design establishes a design basis flood level (including wave action) of one foot below the baseline top of concrete elevation at the ground level floor. Therefore, there are no dynamic flood loads on the RXB and CRB. The lateral hydrostatic pressures on the structures due to the design flood level, as well as ground water and soil pressure, are factored into the structural design as static and dynamic loads discussed in Section 3.8.4.3.3.

2.2 Probable Maximum Precipitation The design utilizes bounding parameters for both rain and snow. The rainfall rate for roof design is 19.4 inches per hour and 6.3 inches for a 5 minute period and the design static roof load because of snow is 50 pounds per square foot. The extreme snow load is 75 pounds per square foot.

The roofs of the RXB and CRB prevent the undesirable buildup of standing water in conformance with Regulatory Guide 1.102 as described below:

The RXB has a gabled roof, with the sloping portions to the north and south. There are no parapets on the top, flat section.

The CRB roof is a sloped steel structure with scuppers in the parapet designed to allow rainfall to drain off the roof. An additional drainage pipe limits the average water depth on the CRB roof to a maximum of four inches.

The bounding rain and snow loads are used in the structural analysis described in Section 3.8.4.

2.3 Interaction of Non-Seismic Category I Structures with Seismic Category I Structures Nearby structures are assessed, or analyzed if necessary, to ensure that there is no credible potential for interactions that could adversely affect the Seismic Category I RXB and CRB. Figure 1.2-2 provides a site plan showing the plant layout. The non-Seismic Category I structures that are adjacent to the Seismic Category I RXB and CRB are:

RWB (Seismic Category II) adjacent to RXB

CRB above elevation 120' (Seismic Category II), above Seismic Category I CRB and adjacent to RXB

North and south Turbine Generator Buildings (Seismic Category III), adjacent to RXB

Central Utilities Building (Seismic Category III), adjacent to CRB

Annex Building (Seismic Category III), adjacent to RXB

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of the probable maximum precipitation.

The RWB has been evaluated and shown to be capable of withstanding the effects of the probable maximum precipitation.

Item 3.4-6: A COL applicant that references the NuScale Power Plant design certification will confirm that nearby structures exposed to external flooding will not collapse and adversely affect the Reactor Building or Seismic Category I portion of the Control Building.

3 References 3.4-1 National Fire Protection Association, "Standard for the Installation of Sprinkler Systems," NFPA 13, 2016 edition, Quincy, MA.

3.4-2 National Fire Protection Association, "Standard for Installation of Standpipe and Hose Systems," NFPA 14, 2016 edition, Quincy, MA.

3.4-3 International Code Council, "Dampproofing and Waterproofing," International Building Code Section 1805, Lenexa, KS, 2015.

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ding Description Pipe Size Flow Isolation Volume of Approximate (in) (gpm) time liquid Volume of liquid (min) (gal) (ft3)

Fire suppression riser 6 2,225 30 66,750 8,900 Fire suppression activities N/A 550 60 33,000 4,400 Chilled water to HVAC 4 226 30 6,780 900 Potable water 2 50 30 1500 200 Fire suppression riser 12 2500 40 100,000 13,400 Fire suppression activities N/A 700 120 84,000 11,200 Main steam 12 77,000 0.0833 6,420 860 Feedwater 8 600 40 24,000 3,200 Site cooling water support for 18 1000 40 40,000 5,400 HVAC Site cooling water header 32 5,000 40 200,000 26,700 Demineralized water 2 300 40 12,000 1,600 Auxiliary boiler 6 80 40 3200 400 CVCS 2-1/2 90 40 3600 500 Pool surge control system 10 2747 40 110,000 14,700 Spent fuel pool/reactor pool 10 --- --- 1,188,600 158,900 cooling 2 3.4-10 Revision 4

Subject to Flood Protection (Without Mitigation)

Building Elevation Room Flood depth (in) Function RXB (( Withheld - See Part 9 010-507 11.25 Mechanical equipment area 010-509 11.25 Mechanical equipment area (( Withheld - See Part 9 }} 010-411 36.75 Steam gallery 010-418 48.0 Steam gallery (( Withheld - See Part 9 }} none (( Withheld - See Part 9 }} 010-207 17.75 Remote shutdown room 010-209 22.75 Battery room 010-210 22.75 Battery room 010-211 22.75 I/O cabinet room 010-212 22.75 Battery room 010-213 22.75 Battery room 010-214 22.75 Battery room 010-215 22.75 Battery room 010-216 22.75 I/O cabinet room 010-217 22.75 Battery room 010-218 22.75 Battery room 010-220 22.75 Battery room 010-221 22.75 Battery room 010-222 22.75 I/O cabinet room 010-223 22.75 Battery room 010-224 22.75 Battery room 010-225 22.75 Battery room 010-226 22.75 Battery room 010-227 22.75 I/O cabinet room 010-228 22.75 Battery room 010-229 22.75 Battery room 010-230 22.75 Battery room 010-231 22.75 Battery room 010-232 22.75 I/O cabinet room 010-233 22.75 Battery room 010-234 22.75 Battery room 010-235 22.75 Battery room 010-236 22.75 Battery room 010-237 22.75 I/O cabinet room 010-238 22.75 Battery room 010-239 22.75 Battery room 010-244 23.25 Battery room 010-245 23.25 Battery room 010-246 23.25 I/O cabinet room 010-247 23.25 Battery room 010-248 23.25 Battery room 010-249 23.25 Battery room 010-250 23.25 Battery room 010-251 23.25 I/O cabinet room 010-252 23.25 Battery room 010-253 23.25 Battery room 010-254 23.25 Battery room 010-255 23.25 Battery room 010-256 23.25 I/O cabinet room 2 3.4-11 Revision 4

Building Elevation Room Flood depth (in) Function 010-257 23.25 Battery room 010-258 23.25 Battery room 010-259 23.25 Battery room 010-260 23.25 Battery room 010-261 23.25 I/O cabinet room 010-262 23.25 Battery room 010-263 23.25 Battery room 010-265 23.25 Battery room 010-266 23.25 Battery room 010-267 23.25 I/O cabinet room 010-268 23.25 Battery room 010-269 23.25 Battery room 010-270 23.25 Battery room 010-271 23.25 Battery room 010-272 23.25 I/O cabinet room 010-273 23.25 Battery room 010-274 23.25 Battery room (( Withheld - See Part 9 }} none 010-107 15.00 Mechanical equipment area (( Withheld - See Part 9 }} 010-114 16.00 Mechanical equipment area 010-125 16.5 Mechanical equipment area 010-134 15.25 Mechanical equipment area (( Withheld - See Part 9 }} none (( Withheld - See Part 9 }} none CRB (( Withheld - See Part 9 }} none (( Withheld - See Part 9 }} none (( Withheld - See Part 9 }} 170-100 17.5 Main control room (( Withheld - See Part 9 }} none (( Withheld - See Part 9 }} none 2 3.4-12 Revision 4

Protection from external missiles is accomplished by locating SSC that require missile protection inside the Seismic Category I Reactor Building (RXB) or Control Building (CRB), or in the Seismic Category II Radioactive Waste Building (RWB). The design complies with General Design Criteria (GDC) 2 and GDC 4 in that structures, systems, and components (SSC) are designed to accommodate the effects of internally and externally generated missiles without losing the ability to perform their safety function. The Seismic Category II RWB is also classified as RW-IIa in accordance with Regulatory Guide (RG) 1.143, "Design Guidance For Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants," Rev. 2. The RWB is designed for the same external missiles as the Seismic Category I structures. This meets or exceeds the design criteria for missiles specified in Table 2 of RG 1.143, Rev. 2. Inside the buildings, missile protection is provided by

providing design features to prevent the generation of missiles.
orienting or physically separating potential missile sources away from equipment subject to missile protection.
providing local shields and barriers for equipment subject to missile protection.

Safety-related SSC and those risk-significant SSC that have a safety function that would be relied upon following the missile producing event are potential missile targets. These structures, systems, and components are located inside the RXB and CRB. Table 3.2-1 lists SSC, their safety classification, and their risk significance. 1 Missile Selection and Description The following potential missile generating sources are considered:

internally generated missiles (outside containment) (Section 3.5.1.1)
internally generated missiles (inside containment) (Section 3.5.1.2)
turbine missiles (Section 3.5.1.3)
missiles generated by tornadoes and extreme winds (Section 3.5.1.4)
site proximity missiles (except aircraft) (Section 3.5.1.5)
aircraft hazards (Section 3.5.1.6)

Missile generation is assumed to occur during all operating conditions. After a potential missile has been identified, its statistical significance is determined in accordance with the following.

1) If the probability of occurrence of the missile (P1) is determined to be less than 10-7 per year, the missile is dismissed from further consideration because it is not statistically significant.

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risk-significant target (P2) is determined. If the combined probability is less than 10-7 per year, the missile and target combination is not considered statistically significant and is dismissed from further consideration.

3) If the product of (P1) and (P2) is greater than 10-7 per year, the probability for damage to the target (P3) is assessed. If the combined probability is less than 10-7 per year, the missile and target combination is not considered statistically significant and is dismissed.

4) If the product of (P1), (P2) and (P3) is greater than 10-7 per year, barriers or other measures are taken to protect the SSC.

1.1 Internally Generated Missiles (Outside Containment) Internally generated missiles are missiles from plant equipment or processes. Missiles can be generated from pressurized systems and components, from rotating equipment, from explosions, or from improperly secured equipment. However, not all potential missiles are credible. The following provides discussion on when missiles do not need to be considered credible (P1 < 10-7). 1.1.1 Pressurized Systems Moderate and low energy systems have insufficient stored energy to generate a missile. As such, the probability of missile occurrence (P1) from systems with operating pressures less than 275 psig is considered to be less than 10-7 (i.e., not credible). Although high energy piping failures could result in dynamic effects, they do not form missiles as such because the whipping section remains attached to the remainder of the pipe. Section 3.6 addresses the dynamic effects associated with pipe breaks. Therefore, potential missiles from high energy piping are the attached components: valves, fasteners, thermowells, and instrumentation. Missiles from piping or valves designed in accordance with ASME Section III, (Reference 3.5-1) and maintained in accordance with an ASME Section XI (Reference 3.5-2) inspection program are not considered credible. Bolted bonnet valves and pressure-seal bonnet valves constructed in accordance with ASME Section III, ASME B16.34, or to an equivalent consensus standard are not considered credible missiles. The use of consensus standards provides reasonable assurance that the components are designed, manufactured and constructed in a manner that demonstrates a high level of quality (e.g., material, design, fabrication, examination, testing). The use of ASME B16.34 and other recognized industry Codes and Standards provides reasonable assurance that the valve maintains its structural integrity during normal and upset conditions and that bolted bonnet valves and pressure-seal bonnet valves cannot become credible missiles. 2 3.5-2 Revision 4

stems with back seats are prevented from becoming missiles by this feature. In addition, the valve stems of valves with power actuators, such as air- or motor-operated valves, are effectively restrained by the valve actuator. Nuts, bolts, nut and bolt combinations, and nut and stud combinations have only a small amount of stored energy and thus are not considered as credible missiles. Thermowells and similar fittings attached to piping or pressurized equipment by welding are not considered as credible missiles. The completed joint has greater design strength than the parent metal. Such a design makes missile formation not credible. Instrumentation such as pressure, level, and flow transmitters and associated piping and tubing are not considered as credible missiles. The quantity of high energy fluid in these instruments is limited and will not result in the generation of missiles. The connecting piping and tubing is made up using welded joints or compression fittings for the tubing. Tubing is small diameter and has only a small amount of stored energy. 1.1.2 Pressurized Cylinders Industrial compressed gas cylinders and tanks are used for the control room habitability system. In addition, smaller portable tanks or bottles used for the chemical and volume control system and maintenance activities may also be stored within the buildings. Cylinders, bottles, or tanks containing highly pressurized gas are considered missile sources unless appropriately secured. The control room habitability system air bottles are mounted in Seismic Category I racks to ensure that each air bottle is contained and does not become a missile. Plates at the end of each bottle retain horizontal movement and pipe straps are installed to prevent vertical movement. Procedures developed in accordance with Section 13.5.2.2 ensure that portable pressurized gas cylinders or bottles are moved to a location where they are not a potential hazard to equipment subject to missile protection, or seismically restrained to prevent them from becoming missiles. 1.1.3 Rotating Equipment The plant design has limited rotating equipment. There are no reactor coolant pumps, turbine driven pumps, or other large rotating components inside the safety-related structures. The main turbine generators are outside of the RXB and are discussed in Section 3.5.1.3. Catastrophic failure of rotating equipment such as fans and compressors leading to the generation of missiles is not considered credible. These components are designed to preclude having sufficient energy to move the masses of their rotating parts through the housings in which they are contained. In addition, material 2 3.5-3 Revision 4

missile generation. 1.1.4 Explosions The battery compartments in the CRB and RXB are ventilated to preclude the possibility of hydrogen accumulation. In addition, the design incorporates valve-regulated lead acid batteries which reduce the hydrogen production in battery rooms compared to vented lead acid batteries. Therefore, a hydrogen explosion in a battery compartment is not a credible missile source. The RWB does not contain any battery compartments. 1.1.5 Gravitational Missiles Structures, systems, and components which could fall and impact or adversely affect safety-related or risk-significant SSC are classified as Seismic Category II (Table 3.2-1). Seismic Category II equipment is mounted to ensure there is no adverse interaction between Seismic Category 1 SSC and Seismic Category II SSC as described in Section 3.2.1.2. These structures, systems, and components are not considered credible missiles. Section 9.1.5 provides an evaluation of the reactor building crane and the module assembly equipment. Due to the significance of a drop of a NuScale Power Module, safety features are designed into these devices as described in Section 9.1.5. Therefore, these devices are not a credible missile source. Procedures developed in accordance with Section 13.5.2.2 ensure that hoisting or lifting activities address movement of heavy loads above safety-related and risk-significant SSC. Control of heavy loads eliminates drops as credible missile sources. Unsecured equipment is a potential gravitational missile. Procedures developed in accordance with Section 13.5.2.2 ensure that maintenance equipment, both equipment brought into the building to perform maintenance, and equipment undergoing maintenance located in the RXB or CRB, are seismically restrained to prevent them from becoming missiles, removed from the building, or moved to a location where they are not a potential hazard. Control of unsecured equipment eliminates falling equipment as credible missile sources. 1.2 Internally Generated Missiles (Inside Containment) There are no credible missiles inside containment. The NPM uses a steel containment that encapsulates the reactor pressure vessel (RPV). There is no rotating equipment inside containment, and all pressurized components are ASME Class 1 or 2 and therefore not credible missile sources as discussed in Section 3.5.1.1.1. 2 3.5-4 Revision 4

core, is non-credible. The CRDM housing is a Class 1 appurtenance per ASME Section III. 1.3 Turbine Missiles The NuScale design employs a barrier approach for protecting essential structures, systems and components (SSC) against the effects of turbine missiles. This approach relies on concrete barriers, in combination with physical separation of redundant safety-related equipment, to meet the requirements of 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 4, Environmental and Dynamic Effects Design Bases, for protecting SSC important to safety. Regulatory Guide (RG) 1.115, refers to SSC important to safety as essential SSC and permits limiting the SSC considered for protection from postulated turbine missiles to those listed in Appendix A of RG 1.115. The Design-Specific Review Standard for the NuScale SMR Design (DSRS) 3.5.1.3, Turbine Missiles, Section I.1 states, Plants that use barriers to protect all essential SSCs specified in RG 1.115, would not have to rely on the turbine missile generation probabilities and turbine rotor integrity discussed in DSRS Section 10.2.3, Turbine Rotor Integrity. The analysis of the missile barriers is performed using the acceptance criteria in Standard Review Plan (SRP) 3.5.3, Barrier Design Procedures. RG 1.115, Revision 2, Section B, Protection against Turbine Missiles, cites shielding as one of the principal means of protecting against turbine missiles. Sections C.2.d and C.3 of RG 1.115 provide guidance for using barriers for protection. NuScale utilized this guidance as the basis for its approach. The acceptance criteria (Section II.6) of DSRS 3.5.1.3, Turbine Missiles, indicates that both high and low trajectory turbine missiles be accounted for. This guidance also appears in RG 1.115. The use of barriers must also meet the acceptance criteria for penetration (and scabbing) of concrete described in RG 1.115 and SRP 3.5.3. The basis for NuScales approach is per RG 1.115. The first component of the safety evaluation is associated with the acceptance criteria when using barriers; no missile can compromise the final barrier, and concrete barriers should be thick enough to prevent backface scabbing (Section C.3 of RG 1.115). The second component of the safety evaluation is associated with conservatively allowing credit for the effect of physical separation of redundant or alternative systems (Section C.1 of RG 1.115). When no backface scabbing occurs, protection is considered adequate. However, when backface scabbing does occur, physical separation of redundant safety-related equipment ensures adequate protection from the bounding turbine missile. The turbine generator building layout in relation to the overall site layout is shown on Figure 1.2-2. As shown in Figure 3.5-1, the turbine generator rotor shafts are physically oriented such that the reactor building (RXB), control building (CRB), and radiological waste building (RWB) are within the turbine trajectory hazard, thereby making the turbines unfavorably oriented with respect to the NuScale Power Modules (NPMs) as defined by RG 1.115, Revision 2. Appendix A of RG 1.115, Rev. 2 identifies essential SSC requiring protection from turbine missiles. For the NuScale design, these SSC are classified as A1 and A2 in Table 3.2-1 and are located in either the RXB or CRB. Table 3.2-1 provides a complete listing of these SSC. Essential SSC within the RXB are protected from turbine missile penetration by the RXB exterior wall. There is some 2 3.5-5 Revision 4

from performing their function. Essential SSC in the CRB are located below grade and are protected by the CRB exterior wall and grade-level slab. 1.3.1 Reactor Building (RXB) The exterior wall of the RXB is the first barrier credited for protecting essential SSC within the building. The reinforced concrete RXB wall penetration shrouds shown in Figure 3.5-2 serve as an additional barrier for equipment located on the 100 ft elevation. However, these shrouds are conservatively not credited in the missile analysis, and the bounding turbine missile is analyzed to impact the exterior wall with no loss of velocity. Detailed description of the RXB exterior walls:

The exterior walls of the RXB are five feet thick. Appendix 3B.2.2.5 provides a description of the exterior wall at grid line E on the south side of the RXB and Figure 3B-23 shows the reinforcement.
Table 3.8.4-10 provides the material properties of the concrete used.
Appendix 3B.2.1 provides the structural material requirements of the RXB.
Section 3.8.4.1.1 provides a description of the RXB with respect to its design category.
Figures 1.2-16 and 1.2-19 show the RXB layout at the grade-level and elevation views.
Table 3.8.4-1 defines the concrete design load combinations.

1.3.2 Control Building (CRB) The grade-level slab and the exterior wall of the CRB are the barriers credited for protecting essential SSC within the building. Essential SSC in the CRB are located below grade. Detailed description of the CRB exterior walls and grade-level slab:

The exterior walls of the CRB are three ft thick. Appendix 3B.3.2.2 provides a description of the exterior wall at grid line 4 on the east side of the CRB and Figure 3B-69 shows the reinforcement.
The grade-level slab of the CRB is three ft thick. Appendix 3B.3.3.2 provides a description of the grade-level slab at elevation (EL) 100'-0".
Table 3.8.4-10 provides the material properties of the concrete used.
Appendix 3B.3.1 provides the structural material requirements of the CRB.
Section 3.8.4.1.2 provides a description of the CRB with respect to its design category.
Figures 1.2-24 and 1.2-26 show the CRB layout at the grade-level and elevation views.

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1.3.3 Bounding Turbine Missile Properties Analyses are performed for three different turbine missiles: a 32.6 pound turbine blade traveling at 1241 mph, a 52.5 pound turbine blade with a fragment of the rotor traveling at 1183 mph, and half of the last stage of the turbine rotor traveling at 476 mph. The turbine rotor is determined to be the most limiting case for the NuScale design. This is based on evaluating the maximum penetration distance for the three postulated missiles and comparing the kinetic energy of the three missiles. The bounding turbine missile is defined as half of the last stage portion of the turbine rotor with the blades attached. It is a semicircular steel section that has a 24 inch radius, weighs 3568 pounds, and travels at 476 mph (based on a 190 percent destructive overspeed of a 3600 rpm turbine). The weight is determined by summing the weight of the semicircular steel rotor portion and 15 turbine blades. The speed is determined by first calculating the distance from the center of the rotor to the centroid of half the rotor with the blades attached, and then multiplying that value by the rotational velocity (and increasing it to 190 percent overspeed). Section 10.2.2 and Table 10.2-1 provide details regarding the type of turbine used in the analysis. 1.3.4 Methodology NuScale assessed the structure of the RXB and CRB for the effects of turbine missiles. The assessment consists of two focus areas: local and global. The local turbine missile barrier assessment evaluates penetration, perforation, and scabbing of the barriers. The global turbine missile barrier analysis evaluates the wall or slab panels of the RXB and CRB to assess the structural response and the ability of the wall or slab to perform its function of resisting design basis loads and transferring loads to adjacent structural elements. The finite element analysis described in the sections that follow, demonstrates that the effects of the bounding turbine missile are local and the wall strain values from the bounding turbine missile are essentially zero at a distance, d, of the missile diameter away from where the impact is applied. As such, for the global assessment, a review of demand-to-capacity ratios from other design basis loads is performed to ensure that adequate margin is available for the bounding turbine missile when applied coincident with those other loads. 2 3.5-7 Revision 4

Acceptance criteria for local damage:

1) Penetration The acceptance criteria for turbine missile barriers as delineated in subsection II.1.A of SRP 3.5.3 suggests the use of empirical equations such as the modified National Defense Research Council (NDRC) formula and A Review of Procedures for the Analysis and Design of Concrete Structures to Resist Missile Impact Effects, by R.P. Kennedy (Reference 3.5-3). Although the methodologies referenced in these documents do not have definitive limits on size or mass, the research and tests that inform the methodology do not use missiles greater than 300 pounds or velocities greater than 500 feet per second. These methods were developed for a steel slug, a piece of steel pipe, and a wooden pole. These objects are relatively small and light when compared to a projectile the size of half of a turbine rotor.

Additionally, the NRDC formula is limited with respect to penetration distance, particularly if the results of the penetration distance divided by the overall thickness are outside of certain values. When this occurs, the NDRC equation results become uncertain. As an alternative, NuScale utilizes a finite element analysis for predicting penetration distance in concrete, instead of the NDRC formula suggested in the acceptance criteria of SRP 3.5.3.

2) Perforation, Spalling, or Scabbing Barrier thickness to protect from perforation, spalling, and scabbing are considered. The modified NDRC formula are used to determine the thickness necessary to prevent perforation and scabbing, based on the penetration thickness determined in the finite element analysis. Spalling is not a concern because it occurs on the face the missile impacts. The required thickness to protect against perforation and scabbing is increased by 20 percent as required by ACI 349-06 Appendix F. The required thickness is then compared to the available thickness of the wall or slab.

Acceptance criteria for global damage: NuScale uses the finite element analysis described in the preceding paragraphs to assess the global effects of the bounding turbine missile. Following the missile impact, the wall or floor needs to serve its function of supporting the structure and transferring forces and moments to adjacent structural elements. To perform this assessment, the concrete and rebar strains are reviewed to determine the extent of damage. Results are then compared to the calculated demand-to-capacity ratios presented in Appendix 3B to confirm that the available margin in the wall or floor is adequate to ensure that each retains its structural integrity and remains capable of performing its safety-related function as described in Part 2 Tier 1, Section 3.11 (RXB) and Section 3.13 (CRB). 2 3.5-8 Revision 4

Concrete barriers require assessment to determine the wall thickness required to prevent barrier perforation and scabbing. First, the concrete barrier penetration depth is determined using FEA. NuScale uses the FEA approach because the penetration depth to missile diameter ratio is outside the limits of the empirical equations commonly used for barrier design. The concrete barrier is modeled using concrete material performance provided through an ANACAP constitutive model. The concrete material properties are modeled to match the RXB or CRB. The compressive strength of concrete used in the analyses is the minimum specified design value. The concrete elastic modulus is calculated from the ACI relation, E=57000fc', and the tensile strength is calculated from the relation ft=1.7*(fc')2/3, where fc' is the effective compressive strength of concrete. Note that these equations use English units (psi). In the analytical concrete material model, the uniaxial cracking strain is used, which is the tensile strength divided by the elastic modulus. The turbine missile is modeled in TeraGrande. The missile material properties, ASTM A470 CL4 (Table 10.2-1), are also modeled. Using actual material properties and barrier and missile geometries, the maximum penetration depth is calculated using nonlinear FEA in TeraGrande. The velocity of the missile is determined by converting the rotational speed of the turbine to a linear velocity of the missile ejected from the turbine rotor. The rotational speed of the NuScale turbine is 3600 rpm and the radius of rotation is taken as the distance from the center of the rotor to the centroid of the missile. TeraGrande performs non-linear analysis factoring in strain hardening of the turbine missile, deformation of the missile, and degradation of the concrete through contact with the missile. The maximum penetration is the distance that the turbine missile penetrates into the concrete barrier before its motion is stopped by the barrier. The wall section considered for each model is 20 ft by 20 ft. A 20 ft by 20 ft section size is used because it is large enough to eliminate any boundary condition effects from the impact of the rotor at the center of the target and serves to assess the global response of the wall to the turbine missile. The floor-to-floor height of the RXB is 26 feet, and an evaluation of the wall strain results of the 20 ft x 20 ft model show that at a distance of one missile diameter away from where the impact is applied, the wall strains approach zero. The above-grade floor-to-floor height of the CRB is 20 ft, which is the same as the height of the wall panel used in the FEA model, and the 20 ft width is shown to adequately capture the behavior of the wall and eliminate any boundary condition effects. The sides of the wall are fixed along all sides parallel to the impact direction. The concrete wall is meshed with 8-node continuum elements. Steel reinforcement is included in the walls and slab (for the CRB), and modeled one-to-one with the layout described in Appendix 3B. The placement of the missile impactor is positioned with the leading edge directly on the center of the target wall. Evaluations also place the rotor strike near the model edge and near a model corner. These evaluations confirm that the 2 3.5-9 Revision 4

the wall. No reduction in velocity is assumed between the location of the turbine and the wall. The rotor fragment is oriented to strike the wall such that it produces the maximum penetration, and is aligned perpendicular to the wall from the sharp edge of the rotor through the center of gravity of the rotor. This produces the largest principal strain and penetration depth as it minimizes the rotation of the rotor on impact with the concrete wall. In addition, investigation is performed with the orientation of the rotor at an alignment that minimizes the projected area of the rotor with the target. 1.3.4.3 Final Required Barrier Thickness Empirical equations endorsed by NUREG 0800 SRP 3.5.3 are used to determine the concrete barrier thickness to prevent perforation of the missile through the barrier and scabbing of concrete material off the back face of the wall. The modified NDRC equations are used. Modified NDRC equations for perforation and scabbing are provided in Section 3.5.3.1.1.2 and Section 3.5.3.1.1.3, respectively. The penetration depth calculated from the FEA is used as input x. The missile diameter is determined from the projected area of the missile on the concrete target. For the RXB, this occurs when the center of gravity of the missile is aligned with the target and results in an equivalent diameter of 22.58 inches. For the CRB, since the missile penetrates the wall and impacts the floor, this occurs when the flat face of the missile impacts the floor, resulting in an equivalent diameter of 27.1 inches. Based on ACI 349 F7.2.1 and F7.2.2, the results from the empirical equations for perforation and scabbing are increased by 20 percent to calculate the final required barrier thickness. The extent of the finite element analyses is a 20 ft x 20 ft wall panel. As such, global damage of the walls is assessed in the same model. An evaluation of reinforcing yielding effects in the wall are evaluated to determine the wall's ability to retain its required capacity. The analysis results show reinforcement yielding is localized to the impact location, which is indicative of a local punching shear mechanism. Reinforcement away from the impact location is seen to be elastic at the end of the impact analyses. As a result, the global assessment focuses on reviewing the overall building results that include the other design basis loads presented in Appendix 3B to ensure that margin is available for the distribution of loads to adjacent elements following the missile perforation. 1.3.4.4 Verification and Validation of Finite Element Analysis Software for Turbine Missiles NuScale performed FEA of the concrete barriers to assess penetration effects of turbine missiles. The software used, TeraGrande Explicit Dynamics Version 2.0-13905, was audited through NuScale's quality assurance program for commercial-grade dedication in accordance with Regulatory Guide 1.231, Acceptance of Commercial-Grade Design and Analysis Computer Programs 2 3.5-10 Revision 4

Plant Engineering: Guideline for the Acceptance of Commercial-Grade Design and Analysis Computer Programs Used in Nuclear Safety-Related Applications, (Reference 3.5-11). To ensure that the FEA software produced penetration results similar to what would be expected of the NuScale turbine missiles, the TeraGrande software was verified and validated against data from EPRI NP-2745, Full-Scale Missile Concrete Impact Experiments for larger missiles (Reference 3.5-12). Table 3.5-2 provides the input parameters for the missiles analyzed. The diameter and shape of the turbine missile and the construction of the target used in the EPRI tests were similar to those of the bounding turbine missile and barriers in the NuScale design. TeraGrande was validated to ensure the software penetration results represent test data published by EPRI for a larger missile. In the EPRI study, case 4 tested a full-scale turbine rotor fragment thrown at a heavily reinforced concrete wall (Table 3.5-2 shows the size and velocity of the missile). This study published results that the 4630 pound turbine rotor fragment penetrated the concrete wall by 25.6 inches and that a steel plate liner on the back face of the wall deformed by 1.77 inches. Parameters of the EPRI test case 4 were modeled using the same rotor missile and concrete wall properties and analyzed using TeraGrande. The FEA results show the rotor missile penetrates the concrete wall 29.6 inches and the steel plate liner on the back face of the wall deforms by 1.81 inches. The FEA results are 15.8 percent higher for penetration and only 2.3 percent higher for deformation of the liner plate. The EPRI test and TeraGrande FEA results demonstrate that the FEA software accurately predicts the penetration depth of a large turbine rotor missile. 1.3.5 Results The RXB is evaluated for both local and global effects of the bounding turbine missile using design basis concrete and rebar material properties. For the local effects, the FEA was used to determine the penetration depth of the bounding missile into the barrier. The FEA results show that the bounding turbine missile at 190 percent overspeed penetrates the 60 inch, reinforced concrete wall a maximum of 51.2 inches. That value, and the minimum projected area, are used as inputs in the NDRC equations to determine the minimum thickness to prevent perforation and scabbing. The results show that some perforation and scabbing occur. However, damage is contained within the gallery space opposite the reactor pool, at either the 100 ft or 125 ft elevations of the RXB. In the event that equipment in either of these rooms is rendered inoperable, the plant remains in a safe condition due to the redundancy of equipment in the other room. Given the physical separation of the redundant safety-related equipment in the NuScale design, no single turbine missile can prevent an essential system from performing its safety function. In addition, the majority of essential equipment that needs to be protected is located on top of the NPM. The NPM is located behind an additional five ft thick concrete reinforced barrier (i.e., the RXB "pool wall") as shown in Figure 3.5-2, and is unaffected by the postulated bounding turbine missile. For the 2 3.5-11 Revision 4

the wall given the turbine missile impact. The maximum principal strains show that, while there is perforation of the wall around the area of the bounding turbine missile application, the overall wall remains intact and serves its function of resisting design basis loads. Specifically, the analysis results show reinforcement yielding is localized to the impact location, which is indicative of a local punching shear mechanism. The analysis also shows that damage is limited to the region where the missile hits, and extends approximately a distance of the missile diameter, d, away from that point. At this location, the principal strain caused by the impact is essentially zero, and as such, the stress in the wall from the missile impact is also zero. This means that the perforation in the wall is limited to a size of the 24 inch radius of the turbine missile. Given that the damage is confined to this local region, it is concluded that the turbine missile has a negligible effect on the overall response of the building. Reinforcement away from the impact location is seen to be elastic at the end of the impact analyses. As a result, the global assessment focuses on reviewing the overall building results that include the other design basis loads as presented in Appendix 3B. Table 3B-14 shows that the maximum element demand-to-capacity ratio for RXB wall reinforcement in any direction is 0.77 (most are below 0.5), which indicates available margin for these and other applicable loads. In other words, the FEA model shows that the wall reinforcing yields only near the point of missile impact, indicating that the stresses on the wall beyond the point of impact are minimal and within the margin of the wall capacity. These analyses, local and global, demonstrate that, while there is perforation from the bounding turbine missile, the wall retains its overall required strength for design basis loads. The control building is also evaluated for local and global effects of the bounding turbine missile using design basis concrete and rebar material properties. Similar to the local analysis of the RXB, a finite element analysis of the CRB is performed. In the analysis, the missile is impacted on the three ft thick reinforced concrete CRB exterior wall (see Figure 3.5-3). As determined by the FEA, the missile at 190 percent overspeed penetrates the CRB exterior wall with a hole that is roughly the size of the missile, and exits with a residual velocity of 223 mph. This residual velocity is then applied at a 15 degree angle to the floor of the CRB and a maximum of 2.9 inches of penetration occurs on the grade-level slab of the building. The angle is determined by conservatively doubling the angle of a direct line from the top of the turbine to the grade-level slab of the building. The thickness to prevent perforation is 10.83 inches and thickness to prevent scabbing is 25.64 inches, both of which are less than the 36 inch thickness of the floor. Therefore, no perforation or scabbing occurs. In general, the missile applies a glancing blow to the floor. Since the safety-related equipment and the main control room are below the grade-level slab, the CRB structure adequately protects against the postulated bounding turbine missile. Similar to the reactor building, the control building finite element analysis is used for the global assessment. The maximum principal strains show that, while the bounding turbine missile penetrates the exterior wall of the CRB, the strain effects are localized to the region of the wall and floor slab where the missile impacts, and they remain intact. Similar to the reactor building, the extent of damage caused by the turbine missile is limited to the size of the missile, and a distance, d, away from the point of impact. The hole in the wall developed 2 3.5-12 Revision 4

exterior wall and grade-level slab, respectively, of the CRB are presented. The maximum demand-to-capacities of the reinforcement in any direction (including shear) are 0.78 (most are below 0.4) and 0.84, demonstrating available margin to these and other applicable loads. In summary, the FEA model shows that the slab reinforcing yields only near the point of missile impact, indicating that the stresses on the slab beyond the point of impact are minimal and within the margin of the slab capacity. The analyses show that the CRB wall and slab adequately retain their structural function so as not to compromise the function of safety-related equipment and the main control room. Evaluation of shock effects from turbine missiles are not explicitly mentioned by SRP 3.5.3 or RG 1.115, but are also considered for turbine missiles. When a missile strikes a barrier, shock is transmitted starting from the center of the initial impact along a structural pathway that may affect essential equipment. The buildings are robust with respect to vibrations, specifically slabs, and anchored equipment are designed to withstand the effects of the in-structure response spectra provided in Section 3.7.2 and will remain functional after the accelerations that are a part of a safe-shutdown earthquake. This provides confidence that the design will also be adequate for shock accelerations that result from a turbine missile impact. In the RXB, systems susceptible to direct missile strike effects are protected from shock accelerations by physical separation; specifically, redundant equipment is separated by different floor elevations and north/south plan dimensions, such that the one train of redundant equipment could be rendered inoperable and the duplicate equipment in a separate room remains available to perform the safety function. System functionality is maintained given this redundancy. In the CRB, safety-related equipment is mounted on the floor of the main control room (or below), and not affected by an impact at the grade-level slab. Given the separation of redundant equipment, shock effects from a turbine missile cannot prevent a system from performing its safety function. Item 3.5-1: A COL applicant that references the NuScale Power Plant design certification will demonstrate that the site-specific turbine missile parameters are bounded by the design certification analysis, or provide a missile analysis using the site-specific turbine generator parameters to demonstrate that barriers adequately protect essential structures, systems, and components from turbine missiles. Parameters to verify are limiting turbine missile spectrum (rotor and blade material properties); turbine rotor design, geometry and number of blades; final design of the reactor building exterior wall; final design of the control building exterior wall and grade-level slab; and location of the turbines with respect to the reactor building and control building. Item 3.5-2: A COL applicant that references the NuScale Power Plant design certification will address the effect of turbine missiles from nearby or co-located facilities. 1.3.6 Regulatory Guide 1.115 Appendix A Essential Systems Regulatory Guide 1.115, Appendix A states that SSC considered for protection from postulated turbine missiles may be limited to the SSC listed in the Appendix and 2 3.5-13 Revision 4

SSC. Some Appendix A SSC do not exist in NuScale's design, others perform no safety-related function, and each is designed to fail to its safe position. For the NuScale design, essential SSC are classified as A1 and A2 in Table 3.2-1 and are located in either the CRB or the RXB. Protection of essential SSC from the postulated bounding turbine missile is achieved as follows:

Essential SSC in the CRB are located below grade and are protected by the CRB exterior wall and grade-level slab.
Essential SSC within the RXB are protected from missile penetration by the RXB exterior wall. There is some perforation and scabbing. However, given the physical separation of the redundant safety-related equipment, there is no missile that can prevent an essential system from performing its function.

Essential SSC listed in Table 3.5-3 and located in the RXB can be grouped into three areas: those located behind the pool wall, those located below grade, and those located generally within the building. 1.3.6.1 Reactor Coolant Pressure Boundary The reactor coolant pressure boundary is located within the RXB behind the exterior wall and behind the pool wall and is adequately protected against the effects of turbine missiles. 1.3.6.2 Main Steam and Feedwater Systems The portions of the main steam and main feedwater systems, up to and including the main steam isolation valves, secondary main steam isolation valves and feedwater isolation valves, are located within the RXB behind the exterior wall and behind the pool wall and are adequately protected against the effects of turbine missiles. 1.3.6.3 The Reactor Core The reactor core is located within the RXB behind the exterior wall and behind the pool wall and is adequately protected against the effects of turbine missiles. 1.3.6.4 Safe Shutdown and Cooling Systems required for attaining safe shutdown are located in the RXB and are protected through redundancy and separation such that the ability of these systems to perform their safety-related functions is not affected by the effects of turbine missiles. There are no SSC required for attaining safe shutdown located outside the RXB. 2 3.5-14 Revision 4

systems to perform their safety-related functions is not affected by the effects of turbine missiles. There are no SSC required for removing residual heat located outside the RXB. The combined capacity of the reactor pool and spent fuel pool (ultimate heat sink) provide the required design basis cooling for spent fuel. As described in Table 1.9-3, the ultimate heat sink is a supply and source for spent fuel cooling for accident conditions. The reactor pool and spent fuel pool are located in the RXB behind the exterior wall and behind the pool wall and are adequately protected against the effects of turbine missiles. The containment system main steam isolation valves and the nonsafety-related secondary main steam isolation valves are located in the RXB behind the pool wall and are adequately protected against the effects of turbine missiles. There are no systems required for supplying makeup water for the primary system to achieve safe shutdown. There are no support systems required for supporting the above functions. 1.3.6.5 Spent Fuel Pool The spent fuel storage pool is located within the RXB behind the exterior wall and behind the pool wall and is adequately protected against the effects of turbine missiles. 1.3.6.6 Reactivity Control Systems As identified in Table 1.9-3, the control rod drive system is the safety-related means of reactivity control for the NuScale design. The control rod drive mechanisms, which de-energize on a loss of power (fail safe), are located in the RXB behind the exterior wall and behind the pool wall and are adequately protected against the effects of turbine missiles. 1.3.6.7 Control Room The control room, including equipment needed to maintain the control room within safe habitability limits for personnel and within safe environmental limits for protected equipment, are located within the CRB and below grade and are adequately protected against the effects of turbine missiles. 1.3.6.8 Gaseous Radwaste Treatment System Portions of the gaseous radwaste treatment system (GRWS) are located in the RXB and in the radiological waste building (RWB). The GRWS components located in the RWB and portions of the RXB are susceptible to the effects of turbine missiles. However, should a portion of the 2 3.5-15 Revision 4

greater than 25 percent of the applicable guideline exposures in 10 CFR 52.47(a)(2)(iv) (equivalent to the guideline exposures of 10 CFR 50.34(a)(1)). Such a release is bounded by the failure of the GRWS postulated in Section 11.3.3.1 and the related doses presented in Table 11.3-9. 1.3.6.9 Monitoring and Actuating Systems The neutron monitoring system (NMS) and the module protection system (MPS) are the systems required for monitoring, actuating, and operating protected portions of the systems listed in Items 4, 6, and 7 of Appendix A of Regulatory Guide 1.115. Item 13 is not applicable to the NuScale design. Portions of the MPS are located in the RXB and the CRB. The RXB houses the NMS equipment. The NMS and MPS components in the RXB are located behind the pool wall or below grade and are protected through redundancy and separation such that the ability of these systems to perform their safety-related functions is not affected by the effects of turbine missiles. The portions of the MPS located within the CRB are located below grade and are adequately protected against the effects of turbine missiles. 1.3.6.10 Electric and Mechanical Devices and Circuitry Electric and mechanical devices and circuitry between the process sensors and the input terminals of the actuator systems involved in generating signals relied on for initiating actions for:

attaining safe shutdown are located in the RXB. These systems are adequately protected through redundancy and separation such that the ability to perform their safety-related functions is not affected by the effects of turbine missiles.
removing residual heat are located in the RXB. These systems are adequately protected through redundancy and separation such that the ability to perform their safety-related functions is not affected by the effects of turbine missiles.
initiating spent fuel cooling. There is no process sensor or actuation system that initiates spent fuel pool cooling. Spent fuel pool cooling system initiation is an operator task. The spent fuel cooling system is nonsafety-related and not risk-significant. The pool volume provides cooling as previously described.
mitigating the consequences of a credible missile-induced high energy line break are located in the RXB and are adequately protected through redundancy and separation such that the ability to perform their safety-related functions is not affected by the effects of turbine missiles.

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reactor coolant system. As described in Section 3.1.4.4, the chemical and volume control system provides reactor coolant make-up during normal operation for small leaks in the reactor coolant pressure boundary, but is not relied upon during a design basis event. Electric and mechanical devices and circuitry between the process sensors and the input terminals of the actuator systems involved in generating signals that are relied on for reactivity control are located in the RXB and the CRB. Those portions located in:

The RXB are adequately protected through redundancy and separation such that the ability to perform their safety-related functions is not affected by the effects of turbine missiles.
The CRB are located below the grade-level slab of the CRB and are adequately protected against the effects of turbine missiles.

There are no electric and mechanical devices and circuitry between the process sensors and the input terminals of the actuator systems involved in generating signals that initiate protective actions from the control room. These components are located in the RXB behind the pool wall under the bioshield or below grade. The requirement to protect electric and mechanical devices and circuitry between the process sensors and the input terminals of the actuator systems involved in generating signals that initiate protective actions by protected portions of the Class 1E electric systems and auxiliary systems for the on-site electric power supplies that provide the emergency electric power needed for the functioning of plant features, is not applicable to the NuScale design. NuScale does not have a Class 1E electric system or any other emergency electric power necessary for the functioning of essential SSC. 1.3.6.11 Long-term Core Cooling Emergency core cooling system equipment and other systems required to provide long-term core cooling (i.e., containment and the ultimate heat sink) after a loss-of-coolant accident are contained within the RXB behind the exterior wall and behind the pool wall and are adequately protected against the effects of turbine missiles. Individual fuel assemblies in the NPM or in the spent fuel pool are located within the RXB behind the exterior wall and behind the pool wall and are adequately protected against the effects of turbine missiles. 1.3.6.12 Containment and Structures The primary reactor containment is located within the RXB behind the exterior wall and behind the pool wall and is adequately protected against the effects of turbine missiles. 2 3.5-17 Revision 4

As described in Tier 1, Section 3.11.1, the RXB is a safety-related building. Protection of safety-related items located in the RXB is achieved through redundancy and separation such that the ability of these systems to perform their safety-related functions is not affected by the effects of turbine missiles. As described in Tier 1, Section 3.13.1, the CRB is a safety-related building. The CRB exterior wall and grade-level slab serve as the barriers to protect safety-related SSC located within the CRB against the effects of turbine missiles. The auxiliary building contains no essential SSC requiring protection from turbine missiles. 1.3.6.13 Electrical Systems NuScale's design has no Class 1E electric systems or auxiliary systems for on-site electric power supplies that provide emergency electric power needed for the functioning of plant features included in Items 1 through 11 of Appendix A of Regulatory Guide 1.115. 1.3.6.14 Other SSC For the NuScale design, there are no SSC or portions of SSC not already identified in items 1 through 13 whose continued function, if lost, could reduce to an unacceptable safety level the functional capability of any plant features included in Items 1 through 13 of RG 1.115, Appendix A. 1.4 Missiles Generated by Tornadoes and Extreme Winds Hurricane and tornado generated missiles are evaluated in the design of safety-related structures and risk-significant SSC outside those structures. The missiles used in the evaluation are assumed to be capable of striking in all directions and conform to the Region I missile spectrums presented in Table 2 of RG 1.76, Rev. 1, "Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants" for tornado missiles and Table 1 and Table 2 of RG 1.221, Rev. 0, "Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants," for hurricane missiles. These spectra are based on the design basis tornado and hurricane defined in Section 3.3.2 and represent probability of exceedance events of 1 x 10-7 per year for most potential sites. The selected missiles include

A massive high-kinetic-energy missile that deforms on impact, such as an automobile.

The "automobile" missile is 16.4 feet by 6.6 feet by 4.3 feet with a weight of 4000 lbs. and a CDA/m (drag coefficient x projected area/mass) of 0.0343 ft2/lb. 2 3.5-18 Revision 4

hurricane. The automobile missile is considered capable of impact at all altitudes less than 30 ft above all grade levels within 1/2 mile of the plant structures.

A rigid missile that tests penetration resistance, such as a six-inch diameter Schedule 40 pipe.

The "pipe" missile is 6.625 inch diameter by 15 feet long with a weight of 287 lbs. and a CDA/m of 0.0212 ft2/lb. This missile has a horizontal velocity of 135 ft/s and a vertical velocity of 91 ft/s in a tornado; and corresponding velocities of 251 ft/s and 85 ft/s, respectively, in a hurricane.

A one-inch diameter solid steel sphere to test the configuration of openings in protective barriers.

The "sphere" missile is 1 inch in diameter with a weight of 0.147 lbs. and a CDA/m of 0.0166 ft2/lb. This missile has a horizontal velocity of 26 ft/s and a vertical velocity of 18 ft/s in a tornado; and corresponding velocities of 225 ft/s and 85 ft/s, respectively, in a hurricane. These missile parameters are key site parameters and are provided in Table 2.0-1. 1.5 Site Proximity Missiles (Except Aircraft) As described in Section 2.2, the NuScale Power Plant certified design does not postulate any hazards from nearby industrial, transportation or military facilities. Therefore, there are no proximity missiles. 1.6 Aircraft Hazards As described in Section 2.2, the NuScale Power Plant certified design does not postulate any hazards from nearby industrial, transportation or military facilities. Therefore, there are no design basis Aircraft Hazards. Discussion of the beyond design basis Aircraft Impact Assessment is provided in Section 19.5. 2 Structures, Systems, and Components to be Protected from External Missiles All safety-related and risk-significant SSC that must be protected from external missiles are located inside the seismic Category I RXB and Seismic Category I portions of the CRB. The concrete walls and roof of the RXB and the CRB below the 30 ft above plant grade threshold are designed to withstand all design basis missiles discussed in Section 3.5.1.3 and Section 3.5.1.4. The portions of the RXB and the CRB that are above 30 ft plant elevation have not been analyzed to withstand the design basis automobile missile, but are 2 3.5-19 Revision 4

Item 3.5-3: A COL applicant that references the NuScale Power Plant design certification will confirm that automobile missiles cannot be generated within a 0.5-mile radius of safety-related structures, systems, and components and risk-significant structures, systems, and components requiring missile protection that would lead to impact higher than 30 feet above plant grade. Additionally, if automobile missiles impact at higher than 30 feet above plant grade, the COL applicant will evaluate and show that the missiles will not compromise safety-related and risk-significant structures, systems, and components. The RXB and CRB meet the requirements of the RG 1.13, Rev. 2, "Spent Fuel Storage Facility Design Basis", RG 1.117, Rev. 2, "Protection Against Extreme Wind Events and Missiles for Nuclear Power Plants," and RG 1.221, Revision 0, "Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants" for protection of SSC from wind, tornado and hurricane missiles. Item 3.5-4: A COL Applicant that references the NuScale Power Plant design certification will evaluate site-specific hazards for external events that may produce more energetic missiles than the design basis missiles defined in Tier 2, Section 3.5.1.4. 3 Barrier Design Procedures In the design, there are a limited number of potential internal missiles and a limited number of targets. If a missile/target combination is determined to be statistically significant (i.e., the product of (P1), (P2) and (P3) is greater than 10-7 per year), barriers are installed. Safety-related and risk-significant SSC are protected from missiles by ensuring the barriers have sufficient thickness to prevent penetration and spalling, perforation, and scabbing that could challenge the SSC. Missile barriers are designed to withstand local and overall effects of missile impact loadings. The barrier design procedures discussed below may be used for both internal and external missiles. 3.1 Local Damage Prediction The prediction of local damage in the impact area depends on the basic material of construction of the structure or barrier (i.e., concrete, steel, or composite). The analysis approach for each basic type of material is presented separately. It is assumed that the missile impacts normal to the plane of the wall on a minimum impact area. 3.1.1 Concrete Barriers Concrete missile barriers are evaluated for the effects of missile impact resulting in penetration, perforation, and scabbing of the concrete using the Modified National Defense Research Committee (NDRC) formulas discussed in "A Review of Procedures for the Analysis and Design of Concrete Structures to Resist Missile Impact Effects," (Reference 3.5-3) as described in the following paragraphs. 2 3.5-20 Revision 4

Concrete thicknesses to preclude perforation or scabbing from the design basis hurricane and tornado pipe and sphere missiles have been calculated for the 5000 psi and 7000 psi concrete used for the RXB, CRB and RWB external walls and roof using the below equations. The design basis hurricane and tornado automobile missile is incapable of producing significant local damage; therefore, it is not considered. The wind and tornado missile results are tabulated in Table 3.5-1. The RXB has five foot thick outer walls and a four foot thick roof. The missile protected portions of the CRB have three foot thick exterior walls and roof, consisting of a concrete slab with a steel cover, and the RWB has exterior walls that are two feet thick above grade and has a one foot thick roof. Additional design characteristics of the RXB and the CRB are provided in Section 3B.2. The RWB exterior walls are 5000 psi concrete reinforced with a minimum of #8 reinforcing bars on 12-inch centers. 3.1.1.1 Penetration and Spalling Equations The depth of missile penetration, x, is calculated using the following formulas: 0.5 V 1.8 x x = 4KNWd --------------- for --- 2.0 Eq. 3.5-1 1000d d V 1.8 x x = KNW --------------- + d for --- 2.0 Eq. 3.5-2 1000d d where, x = penetration depth, in, W = missile weight, lb, d = effective missile diameter, in, N = Missile shape factor:

flat nosed bodies = 0.72,
blunt nosed bodies = 0.84,
average bullet nose (spherical end) = 1.00,
very sharp nosed bodies = 1.14, V = Velocity, ft/sec, K = 180 ( f' c ) , and 2 3.5-21 Revision 4

3.1.1.2 Perforation Equations The relationship for perforation thickness, tp (inches), and penetration depth, x, is determined from the following formulas: 2 t p d = 3.19 ( x d ) - 0.718 ( x d ) for ( x d ) < 1.35 t p d = 1.32 + 1.24 ( x d ) for 1.35 ( x d ) 13.5 3.1.1.3 Scabbing Equations The relationship for scabbing thickness, ts (inches), and penetration depth, x, is determined from the following formulas: 2 t s d = 7.91 ( x d ) - 5.06 ( x d ) for ( x d ) < 0.65 t s d = 2.12 + 1.36 ( x d ) for 0.65 ( x d ) 11.7 3.1.2 Steel Barriers There are no steel missile barriers used in the design. 3.1.3 Composite Barriers The design does not use composite barriers. 3.2 Overall Damage Prediction For predicting overall damage, a dynamic impulse load concentrated at the impact area is determined and applied as a forcing function to determine the structural response. The forcing functions to determine the structural responses are derived using EPRI NP440, "Full Scale Tornado Missile Impact Tests," (Reference 3.5-9) for the triangular impulse formulation of the design basis steel pipe missile. BC-TOP-9A, Rev. 2, "Design of Structures for Missile Impact," (Reference 3.5-8) is used for the design basis automobile missile. The solid sphere missile is too small to affect the structural response of the RXB and the CRB and was not evaluated for its contribution to overall structural response. The automobile missile forcing functions are applied to the building models in selected locations using the horizontal impact loads since they are higher than the vertical loads. The results are addressed in Section 3.8.4. 2 3.5-22 Revision 4

(Reference 3.5-6) for concrete structures and AISC N690 "Specification for Safety-Related Steel Structures for Nuclear Facilities," (Reference 3.5-7) for steel structures except for the modifications listed below. Stress and strain limits for the missile impact equivalent static load comply with applicable codes and RG 1.142, Rev. 2 "Safety-Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments)," and the limits on ductility of steel structures are given as noted below. Concrete Structural concrete members designed to resist missile impact are designed for flexural, shear, spalling, scabbing, and perforation effects using the equivalent static load obtained for the evaluation of structural response. The permissible ductility for beams, walls, and slabs subjected to impulsive or impactive loads, if flexure controls the design, is in accordance with Section F.3.3 of ACI-349. In Section F.3.5 of ACI-349, the permissible ductility ratio (), when a concrete structure is subjected to a pressure pulse due to compartment pressurization, is as follows, based on RG 1.142:

1) for the structure as a whole, 1.0

2) for localized area in the structure (ductility in flexure), 3.0 In Section F.3.7 of ACI-349 where shear controls the design, the permissible ductility ratio is as follows, based on RG 1.142:

1) when shear is carried by concrete alone, 1.0

2) when shear is carried by combination of concrete and stirrups or bent bar, 1.3

3) when shear is carried completely by stirrups, 3.0 In Section F.3.8 of ACI-349, the maximum permissible ductility ratio in flexure is as follows, based on RG 1.142.

1) When the compressive load is greater than 0.1 f'c Ag or one-third of that which would produce balanced conditions, whichever is smaller, the maximum permissible ductility ratio should be 1.0.

2) When the compressive load is less than 0.1 f'c Ag or one-third of that which would produce balanced conditions, whichever is smaller, the permissible ductility ratio should be as given in F.3.3 or F.3.4 of ACI-349.

3) The permissible ductility ratio should vary linearly from 1.0 to that given in F.3.3 or F.3.4 of ACI-349 for condition between specified in 1 and 2.

2 3.5-23 Revision 4

Structural steel members designed to resist missile impact are designed for flexural, shear, buckling and perforation effects using the equivalent static load obtained for the evaluation of structural response. Based on Section NB3.15 of AISC N690, the following ductility factors () from Table NB3.1 are used. 0.25 0.1

1) For steel tension members, ---------------- -------

y y a) u = strain corresponding to elongation at failure (rupture) b) y =strain corresponding to yield stress

2) For structural steel flexural members:

a) Open sections (W, S, WT, etc.), 10 b) Closed sections (pipe, box, etc.), 20 c) Members where shear governs design 5

3) Structural steel columns, = 0.225/(Fy/Fe) st/y (not to exceed 10) a) Fe = 2E/(KLe/r)2 b) Fy = yield strength of steel member c) st = strain corresponding to the onset of strain hardening In determining an appropriate equivalent static load for (Yr), (Yj) and (Ym), elasto-plastic behavior may be assumed with permissible ductility ratios as long as deflections do not result in loss of function of any safety-related system.

Section 3.8 provides additional information on loading combinations and analysis methods for the RXB and CRB. 4 References 3.5-1 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, "Rules for Construction of Nuclear Facility Components," 2013 edition with no Addenda (subject to the conditions specified in paragraph (b)(1) of section 50.55a), Section III, New York, NY. 3.5-2 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2013 2 3.5-24 Revision 4

3.5-3 Kennedy, R.P., "A Review of Procedures for the Analysis and Design of Concrete Structures to Resist Missile Impact Effects," Nuclear Engineering and Designs, (1976) 37:2, 183-203. 3.5-4 Cottrell, W.B., and A.W. Savolainen, "U.S. Reactor Containment Technology," Volume 1, Chapter 6, ORNL NSIC-5, Oak Ridge National Laboratory, Oak Ridge, TN, 1965. 3.5-5 Russel, C.R., Reactor Safeguards, MacMillian, New York, 1962. 3.5-6 American Concrete Institute, "Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary," ACI 349-06, Farmington Hills, MI. 3.5-7 American Institute of Steel Construction, "Specification for Safety-Related Steel Structures for Nuclear Facilities," AISC N690, 2012, Chicago, IL. 3.5-8 Bechtel Power Corporation, "Design of Structures for Missile Impact," BC-TOP-9A, Rev. 2, San Francisco, CA, September 1974. 3.5-9 Electric Power Research Institute, "Full Scale Tornado Missile Impact Tests," EPRI NP440, Palo Alto, CA July 1977. 3.5-10 U.S. Nuclear Regulatory Commission, "Acceptance of Commercial-Grade Design and Analysis Computer Programs Used in Safety-Related Applications for Nuclear Power Plants," Regulatory Guide 1.231, Rev. 0, January 2017. 3.5-11 Electric Power Research Institute, "Plant Engineering: Guideline for the Acceptance of Commercial-Grade Design and Analysis Computer Programs Used in Nuclear Safety-Related Applications," EPRI Report TR-1025243, Rev. 1, December 2013. 3.5-12 Electric Power Research Institute, "Full-Scale Missile Concrete Impact Experiments," EPRI Report NP-2745, February 1983. 2 3.5-25 Revision 4

Concrete Penetration Perforation Thickness Wall/Roof W D V ction Missile N Strength Distance Distance to Preclude Building Thickness (lbs) (in.) (ft/s) (psi) (in.) (in.) Scabbing (in.) 7000 6.2 18.6 23.7 RXB 60 pipe 0.84 287 6.625 251 27.8 from CRB 36 5000 6.7 19.8 EC-F170-3650, Rev 1 RWB 24 ontal 7000 0.3 1.1 2.3 RXB 60 sphere 1.00 0.147 1 224 CRB 36 5000 0.3 1.1 2.4 RWB 24 RXB 48 pipe 0.84 287 6.625 91 5000 2.7 9.4 18.9 CRB 36 RWB 12 tical RXB 48 sphere 1.00 0.147 1 85 5000 0.1 0.5 1.2 CRB 36 RWB 12 2 3.5-26 Revision 4

Missile Type Mass Velocity NP-2745 Test Case 4630 pounds 257 mph cale Turbine Rotor Missile 3568 pounds 476 mph (includes 190% overspeed) 2 3.5-27 Revision 4

sential SSC, or portions thereof, Essential SSC, or portions thereof, Essential SSC, and portions thereof, located inside the RXB located inside the RXB located inside the RXB BEHIND THE POOL WALL BELOW GRADE NOT protected by the pool wall (NOT behind the pool wall) NOT below grade S, Containment System CNTS, Containment System C Injection & Discharge Nozzles

Hydraulic skid C PZR Spray Nozzle C PZR Spray CIV C RPV High Point Degasification zzle C RPV High Point Degasification CIV V & RRV Trip/Reset # 1 & 2 Nozzles V Trip 1 & 2/Reset #3 Nozzles C Injection & Discharge CIVs V Fasteners V Seismic Shear Lug V CRDM Support Frame ntainment Pressure Transducer arrow Range) ntainment Water Level Sensors dar Transceiver) 1 & 2 Steam Temperature Sensors D)

V Close and Open Position Indication S, Supply to SGs and DHR HXs FWIV Steam Generator System tubes egral steam plenums egral steam plenum caps edwater plenums access ports tube supports per and lower SG supports am piping inside containment edwater piping inside containment edwater supply nozzles in steam supply nozzles ermal relief valves edwater plenum access port covers am plenum access ports am plenum access port covers w restrictors Reactor Core System el assembly (RXF) el Assembly Guide Tube S, Control Rod Drive System ntrol Rod Drive Shafts ntrol Rod Drive Latch Mechanism DM Pressure Boundary (Latch using, Rod Travel Housing, Rod vel Housing Plug) Control Rod Assembly components 2 3.5-28 Revision 4

sential SSC, or portions thereof, Essential SSC, or portions thereof, Essential SSC, and portions thereof, located inside the RXB located inside the RXB located inside the RXB BEHIND THE POOL WALL BELOW GRADE NOT protected by the pool wall (NOT behind the pool wall) NOT below grade Reactor Coolant System components (except as listed as B1 B2 in Table 3.2-1) actor vessel internals (upper riser embly (Note 7), lower riser embly, core support assembly, flow erter, and pressurizer spray nozzles) actor vessel internals upper riser llows-lateral seismic restraining ucture de Range RCS Pressure Elements de Range RCS Cold Leg mperature Elements CVCS, Chemical and Volume Control System

DWS Supply Isolation Valves S, Emergency Core Cooling System actor Vent Valve (RVV)

V Trip Valve actor Recirculation Valve (RRV) V Trip Valve set Valve draulic lines S, Decay Heat Removal System Steam Pressure Instrumentation (4 r side) tuation Valve (2 per side) ndenser (1 per side) , Ultimate Heat Sink S Pool (water only; also see RXB and CM below) , Module Protection System MPS, Module Protection System der-the-Bioshield Temperature

All components (except as listed as B1 nsors or B2 in Table 3.2-1)

, Neutron Monitoring System            NMS, Neutron Monitoring System core Neutron Detectors

Excore Separation Group A/B/C/D -

Power Isolation, Conversion and Monitoring Devices

Excore Signal conditioning and processing equipment In-Core Instrumentation System core instrument string sheath [Note:

ovides none of the functions listed in pendix A.] 2 3.5-29 Revision 4

2 3.5-30 Revision 4 Figure 3.5-1: Plan View of Partial NuScale Plant Showing Turbine Missile Trajectory (( Withheld - See Part 9 }} 2 3.5-31 Revision 4

cale Final Safety Analysis Report (( Withheld - See Part 9 }} Missile Protection

This section describes the design bases and measures needed to protect safety-related and essential systems and components inside and outside containment against the effects of postulated pipe rupture. Figure 3.6-1 is a flowchart depicting the steps in the process for evaluation of potential line breaks. The NuScale methodology applicable to identification and assessment of pipe rupture hazards addresses determination of postulated rupture locations, characteristics of ruptures, and assessment of the possible dynamic and external effects of ruptures. Details of the analyses are provided in the Pipe Rupture Hazards Analysis Technical Report (Reference 3.6-21). Pipe rupture protection is provided according to the requirements of 10 CFR 50, Appendix A, General Design Criterion 4. In the event of a high- or moderate-energy pipe rupture within the NuScale Power Module (NPM), adequate protection is provided so that safety-related and essential structures, systems, and components (SSC) are not unacceptably affected. Essential systems and components are those required to shut down the reactor and mitigate the consequences of the postulated piping rupture. Nonsafety-related systems are not required to be protected from the dynamic and environmental effects associated with the postulated rupture of piping except as necessary to preclude adverse effect on an essential system. In addition, although neither safety-related nor essential, the post-accident monitoring (PAM) functionality provided by various portions of the instrumentation and control (I&C) systems is protected. The criteria used to evaluate pipe rupture protection are generally consistent with NRC guidelines including those in the Standard Review Plan Section 3.6.1, Section 3.6.2, and Section 3.6.3, NUREG-1061, Vol. 3, and applicable Branch Technical Positions (BTPs), as discussed within this section. Section 3.6.1 identifies the high- and moderate-energy lines that have a potential to affect safety-related and essential SSC, and describes the approaches used in the NuScale Power Plant design for protection of these SSC. Section 3.6.2 describes the analytical methodology used to determine break locations, identifies postulated breaks, and discusses the consequences of those breaks and the effect on SSC functionality. Section 3.6.3 describes the leak-before-break (LBB) analysis for applicable piping systems inside containment. Section 3.6.4 discusses the analysis of non-LBB high- and moderate-energy piping. 1 Plant Design for Protection against Postulated Piping Ruptures in Fluid Systems General Design Criterion (GDC) 4 requires that SSC be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents (LOCAs). This includes both environmental effects (temperature changes, pressure changes, humidity changes, and flooding) from line breaks and leakage cracks and dynamic effects (blast, pressurization, pipe whip and jet impingement) that may result from high-energy line breaks (HELB). Plant designers are provided with options to address GDC 4 for pipe ruptures. These options are as follows: 2 3.6-1 Revision 4

fatigue limits, and a high level of inservice inspection (ISI). The criteria for this exclusion are provided in BTP 3-4, Fluid System Piping in Containment Penetration Areas, Section B.A.(ii).

Systems that can demonstrate a low probability of rupture prior to the detection of a leak may be excluded from HELB dynamic effect considerations. This is referred to as LBB analysis and is discussed in SRP 3.6.3. LBB is applied to high-energy piping systems having well-characterized loading conditions and load combinations. This method is an acceptable design approach provided that plant design and specific analyses have indicated a low probability of rupture from damage mechanisms such as water hammer, steam hammer, stress corrosion cracking, and fatigue.
For high- and moderate-energy systems that cannot be fully excluded using criteria of BTP 3-4 Section B.A.(ii) or LBB, line breaks and leakage cracks are postulated. The criteria for the specific locations for the postulated breaks are provided in BTP 3-4 (e.g.,

Section B.A.(iii)). In general, locations meeting certain stress, fatigue and design requirements may be excluded and are not required to be postulated to rupture. Other locations, such as terminal ends or high-stress locations, must be postulated to rupture. At break locations, the piping systems are located such that there is no safety-related or essential equipment in the area (i.e., separation), or safety-related and essential SSC are shown to be protected from exposure to break effects or otherwise not unacceptably affected. The piping systems that must be considered include the ASME Section III Class 1, 2, 3, and ASME B31.1, high-energy and moderate-energy systems located inside and outside of the containment vessel (CNV). Table 3.6-1 identifies the high- and moderate-energy piping systems and associated plant locations. Breaks and leakage cracks need not be postulated in high- and moderate-energy lines that are NPS 1 and smaller. High-energy lines are evaluated for both line breaks and through-wall leakage cracks. Line breaks include both circumferential (complete rupture around the circumference of the pipe) and longitudinal breaks (rupture of the pipe along its axis). Line breaks are analyzed for dynamic and environmental effects. Through-wall leakage cracks are analyzed for flooding and environmental effects. Moderate-energy lines are evaluated for through-wall leakage cracks and analyzed for flooding and environmental effects. Additionally, the environmental effects of nonmechanistic breaks of main steam system (MSS) and feedwater system (FWS) piping in the containment penetration area are evaluated. Locations having the greatest effect on essential equipment are chosen for evaluation of impacts. Flooding is discussed in Section 3.4. Environmental effects are discussed in Section 3.11. Analysis of subcompartment pressurization effects within the CNV are discussed in Appendix 3A. 2 3.6-2 Revision 4

High-energy fluid systems include those systems or portions of systems where either of the following conditions is met:

the maximum operating temperature exceeds 200 degrees F, or
the maximum operating pressure exceeds 275 psig Moderate-energy fluid systems include systems or portions of systems pressurized above atmospheric pressure during normal plant conditions but do not meet the criteria for high-energy systems. Moderate-energy fluid systems are those systems where both of the following conditions are met: (a) the maximum operating temperature is 200 degrees F or less, and (b) the maximum operating pressure is 275 psig or less. In addition, piping systems that exceed 200 degrees F or 275 psig for 2 percent or less of the time during which the system is in operation or that experience high-energy pressures or temperatures for less than 1 percent of the plant operation time are also considered moderate-energy.

Table 3.6-1 provides a list of high- and moderate-energy piping systems and identifies the areas where the systems are located. The areas of the plant that contain high- and moderate-energy lines, or safety-related and essential SSC are considered in six groups. Each is discussed in a separate section.

inside the CNV (Section 3.6.1.1.1)
outside the CNV (under the bioshield) (Section 3.6.1.1.2)
in the Reactor Building (RXB), (outside the bioshield) (Section 3.6.1.1.3)
in the Control Building (CRB) (Section 3.6.1.1.4)
in the Radioactive Waste Building (RWB) (Section 3.6.1.1.5)
onsite (outside the buildings) (Section 3.6.1.1.6)

Table 3.6-1 identifies the largest piping line size and the highest normal operating pressure and temperature of the fluid system to assign an energy classification. The energy classification and line size do not necessarily correspond to the same region of the fluid system. While Table 3.6-1 provides a listing of the high- and moderate-energy systems outside of the NPM, the piping line size and energy classification may vary from these maximum values at the postulated rupture location. COL Item 3.6-1 requires that the COL applicant confirm the content of Table 3.6-1 following the performance of the balance of plant PRHA, or update it accordingly. Figure 6.6-1 shows the high- and moderate-energy lines that interface with the CNV. Generally, the portions of these lines from the NPM disconnect flanges up to and including the CNV penetration are considered to be part of the containment system (CNTS). Inside the CNV, the lines are considered to be part of a different system. The main steam and feedwater lines are part of the steam generator system (SGS) inside containment. The chemical and volume control system (CVCS) lines are part of the reactor coolant system (RCS) inside the CNV, and include the RCS injection, RCS 2 3.6-3 Revision 4

return lines are part of the control rod drive system (CRDS) inside the CNV. The decay heat removal system (DHRS) piping is a high-energy system only associated with the NPM. The containment flooding and drain system (CFDS) is a single open pipe inside containment that is normally isolated and not pressurized during operation. This line is moderate-energy based on the amount of time in use. This line is identified as the CNTS flooding and drain line both inside and outside the CNV. Generally, in this Section a particular portion of piping is referred to by its functional name (e.g., main steam, RCCWS) regardless of whether that portion is inside the CNV, a part of the CNTS, or outside the NPM. 1.1.1 Inside the Containment Vessel The high-energy lines inside the CNV are: main steam, feedwater, RCS injection, RCS discharge, high point degasification, PZR spray supply and DHRS condensate return. There are two moderate-energy lines inside the CNV, the RCCWS supply and return lines and the CFDS line (See Table 3.6-1). The ECCS includes several small hydraulic lines inside containment that run between the ECCS valves, the Trip/Reset valves, and the RCS injection line. These high-energy ECCS lines are excluded from consideration as they are smaller than NPS 1. 1.1.2 Outside the Containment Vessel (Under the Bioshield) The high-energy lines (main steam, feedwater, RCS injection, RCS discharge, high point degasification, PZR spray supply and DHRS) and the moderate-energy lines (CRDS, CFDS, and the containment evacuation system (CES)) continue outside containment to the NPM disconnect flange (See Table 3.6-1). The DHRS steam line connects to the MSS line outside containment, immediately upstream of the MSS containment isolation valve and leads to the DHRS condenser and then to the DHRS condensate return lines. Although not normally in use, this entire system is pressurized during NPM operation. 1.1.3 In the Reactor Building (Outside the Bioshield) Within the RXB, but outside the area under the bioshield, the high-energy lines include the MSS, FWS, and CVCS lines, and additional high-energy lines associated with the auxiliary boiler and process sampling system (PSS) (See Table 3.6-1). Based on limited operating time, the auxiliary boiler lines are considered moderate-energy. Based on the nominal diameter of the PSS lines, breaks do not need to be postulated. The high-energy MSS and FWS lines exit the reactor pool through the North and South reactor pool walls, cross a mechanical equipment area (pipe gallery) and exit the RXB. 2 3.6-4 Revision 4

50' 0 and associated CVCS rooms at Elevations 24 0 and 35' 6". The pipe chase can be seen on the general arrangement drawings in Section 1.2. Moderate-energy lines are routed throughout the RXB (See Table 3.6-1). 1.1.4 In the Control Building There are no high-energy lines in the CRB. There are three moderate-energy lines: fire protection, chilled water, and potable water (See Table 3.6-1). 1.1.5 In the Radioactive Waste Building There are no high-energy lines in the RWB. There are two moderate-energy lines: fire protection and liquid radioactive waste management (See Table 3.6-1). 1.1.6 Onsite (outside the buildings) Outside of the RXB and CRB there are three high-energy lines: MSS, FWS, and extraction steam, and multiple moderate-energy lines (See Table 3.6-1). There is no safety-related or essential equipment in the area outside of the RXB or CRB. Final routing of piping outside of the RXB, CRB, and RWB is the responsibility of the COL applicant. Item 3.6-1: A COL applicant that references the NuScale Power Plant design certification will complete the routing of piping systems outside of the containment vessel and the area under the bioshield, identify the location of high- and moderate-energy lines, and update Table 3.6-1 as necessary. This activity includes the performance of associated final piping stress analyses, design and qualification of associated piping supports, evaluation of subcompartment pressurization effects (if applicable), and completion of the Balance of Plant Pipe Rupture Hazards Analysis, including the design and evaluation of pipe whip/jet impingement mitigation devices as required. This includes an evaluation and disposition of multi-module impacts in common pipe galleries. 1.2 Identification of Safety-Related and Essential Structures, Systems, and Components By design, the NuScale Power Plant only has a small number of safety-related and essential SSC. These SSC are primarily associated with the NPM, either inside the CNV or mounted on the top of the CNV head. Shutdown of the reactor requires the following systems be protected from HELB:

RCS
module protection system (MPS)
neutron monitoring system 2 3.6-5 Revision 4
CVCS
control rod assembly and the CRDS
CNTS
DHRS
emergency core cooling system (ECCS)
ultimate heat sink / reactor pool Of these, only the CNTS, DHRS, ECCS, and ultimate heat sink/reactor pool are needed following reactor shutdown. In addition, PAM functionality for Type B and C variables (there are no Type A variables) is protected.

1.3 Characteristics of the NuScale Design The NuScale design is an integral, multi-unit, small modular reactor for which safety is provided by passive features without the need for safety-related electrical power. Because NRC regulatory guidance for HELB is premised on the existing fleet of large light water reactors with reactor coolant loops and active safety features, instances exist where the current NRC pipe rupture guidance is not a direct fit. In many cases, the NRC has not issued a Design-Specific Review Standard to address what is directly applicable for the NuScale design. Specific examples of relevant design differences are:

The response to HELB for a NuScale plant requires neither electric power nor injection of additional cooling water.
The NPMs are mostly submerged in a large pool of water that serves as the ultimate heat sink and does not require replenishment for design-basis events.
Design-basis accidents do not require operator actions or re-establishing electric power for long-term cooling.
Piping is small compared to the large reactors for which regulatory guidance was initially developed.
Active safety-related components (e.g., ECCS valves, DHRS actuation valves, and containment isolation valves (CIVs)) are shown to operate during refueling. As part of the start-up sequence for an NPM, each of the safety-related ECCS, DHRS, and containment isolation valves is repositioned. These system line-up activities provide assurance the safety-related valves are operable.
The NPM containment is a pressure vessel designed and constructed to ASME Code Section III Class 1 requirements versus a building in conventional LWRs.
Piping of the NPM, including secondary system piping, is made of corrosion-resistant stainless steel.
MSS and FWS piping inside the containment boundary and under the bioshield is designed to RCS design pressure and temperature.
MSS and FWS piping inside the CNV meets LBB criteria.

2 3.6-6 Revision 4

The length of piping in which breaks must be postulated is minimal and the size of high-energy piping is small compared to current design reactors.
The NPM containment is operated at a vacuum.
Equipment and piping inside the NPM containment are not covered by insulation.

This is important for multiple reasons: Jet impingement does not dislodge insulation that could lead to blockage of long-term-cooling recirculation. Detection of small leakage cracks is not impeded by retention of moisture in insulation. The bare piping is readily inspectable during refueling, because insulation does not need to be removed to observe deposits, discoloration, or other signs of degradation. Corrosive substances (e.g., chlorides) cannot be trapped and held in contact with the piping surface.

Safety-related and essential components inside the NPM containment are qualified to be functional after exposure to saturated steam at containment design pressure of at least 1000 psia, requiring designs that are robust.
The small NPM containment results in congestion that makes difficult the addition of traditional piping restraints and the separation of essential components from break locations, but whipping pipes in turn have a limited range of motion before encountering an obstacle.
Containment isolation valves are outside of containment. Where two valves in series are required (i.e., for containment penetrations governed by GDC 55 and 56),

both are in a single-piece valve body (i.e., no piping or welds between CIVs, precluding breaks in between). Also, the lines directly connected to the primary system or the containment have only a single piping weld in the area between the containment wall and the CIV.

The RCS-connected lines (i.e., CVCS), except for the normally isolated RPV high-point degasification line, have check (or excess flow check) valves immediately outside the CIVs to preserve reactor coolant inventory in case of LOCAs outside containment.
Containment pressure suppression is not required, and there are no sprays that introduce chemical additives.
During a refueling, the NPM is disconnected from supporting systems by removal of piping spools, transported by crane to a refueling location, and disassembled.

This provides access for inspection to portions of the plant not normally accessible.

Up to 12 NPMs are operating at the same time and in proximity, so the potential for a rupture in a system of one module to affect others is considered.
The plant main control room is in a separate building that does not contain high energy piping systems.

2 3.6-7 Revision 4

different compartments within the building. These unique characteristics affect choices about the means to address HELB. 2 Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping This section describes the criteria and methods used to postulate break and leakage crack locations in high-energy and moderate-energy piping inside and outside containment, the methodology used to define potential blast effects, the thrust at the postulated break location, potential pipe whip, the jet impingement loading on adjacent essential safety-related SSC and subcompartment pressurization resulting from fluid blowdown. General Design Criterion 4 requires that SSC important to safety both accommodate the effects of, and are compatible with, the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents. In the event of a high-energy or moderate-energy pipe rupture within the plant, GDC 4 requires that adequate protection is provided so that essential SSC are not impacted unacceptably by the adverse effects of the rupture. Nonsafety-related systems are not required to be protected from the dynamic and environmental effects associated with the postulated rupture of piping. Compliance with GDC 4 is demonstrated through conformance with the criteria of BTP 3-4 as described in Section 3.6.2.1. 2.1 Criteria Used to Define Break and Crack Location and Configuration Branch Technical Position 3-4 provides guidance on the selection of the break locations within a piping system. The types of breaks postulated in high-energy lines include circumferential breaks in fluid system piping greater than 1 inch NPS; longitudinal breaks in fluid system piping that is 4-inch NPS and greater, and leakage cracks in fluid system piping greater than 1-inch NPS. Leakage cracks are also postulated in moderate-energy lines. The pipe break criteria of BTP 3-4 include the requirement that breaks be postulated at terminal ends. The definition of a terminal end, consistent with BTP 3-4, is the extremity of a piping run that connects to structures, components (e.g., vessels, pumps, valves), or pipe anchors that act as rigid constraints to piping motion and thermal expansion. A branch connection on a main piping run is a terminal end for the branch run, except where the branch run is classified as part of a main run in the stress analysis or is shown to have a significant effect on the main run behavior. In piping runs that are maintained pressurized during normal plant conditions for a portion of the run (i.e., up to the first normally closed valve), a terminal end is the piping connection to this closed valve. General Design Criterion 4 allows dynamic effects associated with postulated pipe ruptures to be excluded from the design basis when analyses demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping. This is referred to as LBB analyses. This is discussed in Section 3.6.3. Similarly, breaks and leakage cracks may be excluded within the containment penetration area if criteria of BTP 3-4 B.A.(ii) are met. 2 3.6-8 Revision 4

The CIVs are outside the containment. A break inside the CNV does not lead to containment bypass. Therefore, there is no containment penetration area inside the CNV and BTP 3-4 B.A.(ii) does not apply. Due to the tight configuration and the concentration of safety-related and essential SSC inside the CNV, dynamic effects of pipe breaks are assessed for specific locations. The following strategies are employed for HELB inside containment:

The main steam and feedwater lines meet the criteria for LBB (see Section 3.6.3). Therefore, circumferential and longitudinal breaks are not postulated for dynamic effects for the MSS and FWS lines inside containment.
The RCS injection, RCS discharge, PZR spray supply, and high-point degasification lines inside containment are NPS 2, Schedule 160, ASME Class 1 stainless steel pipes. Due to their size, longitudinal breaks are not postulated.

Circumferential breaks are postulated in accordance with BTP 3-4 Section B.A.(iii)(1). Breaks in Class 1 high-energy piping systems are postulated at the following locations: a) terminal ends (defined in Section 3.6.2.1) b) intermediate locations where the maximum stress range exceeds 2.4 Sm as calculated by equation (10) and either equation (12) or (13) of NB-3653 of Section III of the ASME Boiler and Pressure Vessel Code. c) intermediate locations where the cumulative usage factor exceeds 0.1, unless environmentally assisted fatigue is considered in which case the cumulative usage factor exceeds 0.4.

The DHRS condensate lines inside containment run from each feedwater line, just upstream of the feed plenum, to the containment upper cylindrical shell penetration. These lines are NPS 2 ASME Class 2. Due to their size, longitudinal breaks are not postulated. Circumferential breaks are postulated in accordance with BTP 3-4 Section B.A.(iii)(2). Breaks in Class 2 high energy piping systems are postulated at the following locations:

a) terminal ends (defined in Section 3.6.2.1) b) at intermediate locations where stresses are calculated by the sum of equations (9) and (10) in NC-3653 of Section III of the ASME Boiler and Pressure Vessel Code to exceed 0.8 times the sum of the stress limits given in NC/ND-3653. The RCCWS and CFDS lines are moderate-energy. Moderate-energy lines are subject only to through-wall leakage cracks and the resultant environmental consequences of localized flooding and increased temperature, pressure, and humidity (Section 3.6.1.2). The environmental effects of postulated moderate-energy leakage cracks are bounded by the accident conditions inside the CNV. As a result, leakage cracks are not evaluated further for the RCCWS and CFDS lines inside containment. 2 3.6-9 Revision 4

ITAAC A07, Pipe Break Hazards Protective Features Verification, was established to confirm that the final pipe rupture hazards analysis demonstrates the acceptability of the dynamic and environmental effects associated with postulated ruptures in high-energy and moderate-energy piping systems within the NPM. 2.1.2 Pipe Breaks Outside the Containment Vessel (under the bioshield) The CIVs for the RCS injection, RCS discharge, PZR spray supply, and RPV high-point degasification lines are each dual, independent valves in a single body that is welded directly to an Alloy 690 safe-end that is welded to the respective nozzle on the CNV head. These lines, except for the normally isolated RPV high-point degasification line, also have a check (injection and spray) or excess flow check (discharge) valve welded directly to the CIV. The feedwater system CIV is similar, except there is a single isolation valve (in accordance with GDC 57) with a check valve as the outboard valve in the single piece body. The MSS lines each have a single CIV. Between the CNV safe end and the valve body are two tee fittings to which the DHRS steam lines attach. Outboard of the valves in each of these lines is a short piping segment welded to a flange used to connect the refueling pipe spools to the module. The containment isolation valves are outside the containment. The containment penetration area is defined by regulatory guidance as the run of piping from the inside CIV to the outside CIV. This definition is not directly applicable to NuScale. Instead, NuScale has omitted piping inside the CNV, but includes the above described valves. In other words, the NuScale containment penetration area is limited to the section from the CNV safe-end-to-valve (or tee) weld out to and including the piping weld to the outermost of the CIV or check/excess flow check valve. For welds in the containment penetration area, provisions of BTP 3-4 Section B.A.(ii) have been applied to preclude the need for breaks to be postulated, because they meet the design criteria of the Section III of the ASME Boiler and Pressure Vessel Code, Subarticle NE-1120 and the following seven criteria:

1) The ASME Class 1 piping (i.e., the four CVCS lines) is designed to satisfy the following stress and fatigue limits:

a) The maximum stress range between any two load sets (including the zero load set) calculated by equation (10) in Section III of the ASME Boiler and Pressure Vessel Code, NB-3653 does not exceed 2.4 Sm. Or, if the calculated maximum stress range of equation (10) exceeds 2.4 Sm, the stress ranges calculated by both equation (12) and equation (13) in Section III of the ASME Boiler and Pressure Vessel Code, NB-3653 meet the limit of 2.4 Sm. 2 3.6-10 Revision 4

0.4. c) The maximum stress, as calculated by equation (9) in Section III of the ASME Boiler and Pressure Vessel Code, NB-3652 under the loadings resulting from a postulated piping rupture beyond these portions of piping, does not exceed 2.25 Sm and 1.8 Sy. The ASME Class 2 main steam and feedwater piping from the safe end to the weld outboard of the body holding the CIV and check valve is designed to satisfy the following stress limits: a) The maximum stress ranges as calculated by the sum of equations (9) and (10) in Paragraph NC-3653, Section III of the ASME Boiler and Pressure Vessel Code, do not exceed 0.8(1.8 Sh + SA). b) The maximum stress, as calculated by Section III of the ASME Boiler and Pressure Vessel Code, paragraph NC-3653 equation (9) under the loadings resulting from a postulated piping rupture of fluid system piping beyond these portions of piping, does not exceed 2.25 Sh and 1.8 Sy.

2) There are no welded attachments for pipe supports.

3) There is a minimum number of circumferential and no longitudinal welds in these lines in the containment penetration area.

4) The length of the piping is the minimum practical (the total containment penetration piping length for 12 NPMs is less than a typical large pressurized water reactor).

5) There are no pipe anchors or restraints.

6) Guard pipes are not used.

7) The piping welds are included in the ISI program as described in Section 6.6, and the NPS 2 welds including and inboard of those of the pipe to outer nozzle welds of the check and excess flow check valves and CIVs are 100 percent volumetrically inspected, in addition to surface inspections as required by the ASME Boiler and Pressure Vessel Code Section XI.

Outboard of the containment isolation valves and check/excess flow check valves, the CVCS NPS 2, Schedule 160, RCS discharge, RCS injection, PZR spray supply, and high point degasification lines are ASME Class 3 lines to the first spool piece used to disconnect the NPM from the permanent piping. The spool piece and subsequent piping are also ASME Class 3 to the junction of an additional valve (or check valve) in each line, and subsequently become ASME B31.1 after that last valve. At the first spool piece breakaway flange, the four lines become part of the CVCS. Remaining piping under the bioshield, including the refueling pipe spools, is designed to comply with BTP 3-4 Rev. 2 Paragraph B.A.(iii) to preclude breaks at intermediate 2 3.6-11 Revision 4

exceed 0.8 times the sum of the stress limits given in NC/ ND-3653. Final stress analysis is performed concurrent with fabrication of the first NPM. Based on designing to meet the criteria from BTP 3-4, no breaks in the NPM bay outside the CNV (under the bioshield) are postulated. However, nonmechanistic breaks in MSS and FWS lines in the containment penetration area and leakage cracks are considered. Decay Heat Removal System Lines The DHRS is a closed loop system outside of the CNV that is entirely associated with a single NPM. Each NPM has two independent DHRS trains. Each train is associated with an independent steam generator (SG). The only active components in the DHRS are the DHRS actuation valves. The DHRS also relies on the MSS and FWS containment isolation valves to provide a closed loop system when it is activated. The DHRS is used to respond to transients including HELB outside containment. It is not used for normal shutdown, though the DHRS actuation valves are opened to allow slight circulation during wet layup of the SG. There is no flow through the DHRS system during normal operation. The DHRS is attached to the MSS line between the CNV and the MSS CIV. This portion of DHRS has two parallel actuation valves that are normally closed. These two lines join into a single line that supplies the passive condenser. Each DHRS condenser is attached to the outside of the CNV. The condenser is designed as an ASME Class 2 component. A NPS 2 line exits the bottom of each DHRS condenser and penetrates the CNV. This line connects to the feedwater system inside containment. During operation, the DHRS is pressurized from the feedwater line. See Section 5.4.3 for additional discussion about the DHRS. Breaks are not postulated in the DHRS piping outside containment in accordance with in BTP 3-4, B.A.(ii). Subject to certain design provisions, NRC guidance allows breaks associated with high-energy fluid systems piping in containment penetration areas to be excluded from the design basis. Though the DHRS piping extends beyond what would traditionally be considered a containment penetration area, this approach is chosen because the DHRS cannot be isolated from the CNV as there are no isolation valves. Breaks are not postulated in this segment of piping because it meets the design criteria for break exclusion in a containment penetration area (see Section 3.6.2.1.2). Although the DHRS condenser is manufactured from piping products, it is considered a major component and not a piping system; thus breaks are not postulated. Item 3.6-2: A COL applicant that references the NuScale Power Plant design certification will verify that the pipe rupture hazards analysis (including dynamic and environmental effects) of the high- and moderate-energy lines outside the containment vessel (under the bioshield) is applicable. If changes are required, the COL applicant will update the pipe rupture hazards analysis, design additional protection features as necessary, and update Table 3.6-2. 2 3.6-12 Revision 4

BTP 3-3 B.1 (a)(1) specifies:

 "Even though portions of the main steam and feedwater lines meet the break exclusion requirements of item 2.A(ii) of BTP 3-4, they should be separated from essential equipment. Designers are cautioned to avoid concentrating essential equipment in the break exclusion zone. Essential equipment must be protected from the environmental effects of an assumed non-mechanistic longitudinal break of the main steam and feedwater lines. Each assumed non-mechanistic longitudinal break should have a cross sectional area of at least one square foot and should be postulated to occur at a location that has the greatest effect on essential equipment."

For the NuScale design, the following considerations apply:

MSS and FWS piping is the largest, high energy piping near the containment boundary
The lines have a single CIV outside containment in accordance with GDC 57 for lines closed inside containment
MSS and FWS piping is usually made of less corrosion resistant material than used for the NuScale design. MSS and FWS piping in many pressurized water reactors is carbon or low alloy steel, which has greater susceptibility to degradation than stainless steel.

Analyzing for non-mechanistic ruptures provides assurance that multiple essential SSCs are capable of withstanding the effects of a limited piping failure should one occur. In the NuScale plant, the dual CIVs are located outside the containment and exposed to the same environmental conditions, which makes protection against unexpected ruptures particularly important. However, the NuScale design has the following characteristics that make non-mechanistic ruptures low risk:

The essential SSCs in vicinity of MSS and FWS piping in the containment penetration area are CIVs, DHRS valves, and instrumentation cables and sensors.
Unlike some safety-related valves in other plant designs that use motor-operators, the NuScale CIVs are hydraulically held open against pneumatic pressure from an accumulator and shut upon a loss of power or a failure of the hydraulic line. The DHRS actuation valves similarly fail open.
Failure of MSS and FWS piping is unlikely because:

Piping in the containment penetration area is made of stainless steel. The physical length of MSS and FWS piping in the containment penetration area is zero (i.e., there are only valves and fittings). MSS and FWS piping has a design pressure and temperature of 2100 psia and 625 degrees F, respectively, equivalent to the RCS piping. 2 3.6-13 Revision 4

break is disproportionately large for a small modular reactor with small pipe sizes. NuScale MSS piping is NPS 12 Schedule 120 and FWS piping is NPS 4 and NPS 5 Schedule 120 in the containment penetration area. For those piping sizes, a 1 ft2 flow area exceeds the area for a full circumferential rupture, which is physically unrealistic. For the NuScale design, non-mechanistic breaks of MSS and FWS piping in the containment penetration area are evaluated, after consideration of the design differences from larger LWR plants. Comparing the typical PWR pipe MSS flow area to that of NuScale (NPS 30 to 38 vs NPS 12) yields a ratio of one-eighth to one twelfth. On this basis, NuScale analyzes for environmental effects of an MSS non-mechanistic break with an area of 12 in2, versus 1 ft2 (144 in2). The non-mechanistic FWS break size applied for the NuScale design (NPS 4 and NPS 5) is 5.87 in2. The volume under the bioshield is small; roughly a cube 20 feet on a side. Therefore, even though only leakage cracks are required to be considered outside the containment penetration area, analysis is performed for a 12 in2 MSS break at the highest point of the pipe run, resulting in a conservative pressure and temperature profile over time for environmental qualification and bounding breaks occurring in any section of the piping under the bioshield. 2.1.2.2 Break Exclusion BTP 3-4 B.A.(iii) identifies specific criteria for which ruptures need not be considered from the containment wall to and including the inboard or outboard isolation valves (usually referred to as the containment penetration area "break exclusion zone"). The concept was necessary due to constraints on ability to cope with breaks between the CIVs. Should a break occur between the CIVs followed by a single failure of a CIV, then containment bypass could occur. To preclude bypass, criteria were developed to ensure that the probability of a piping failure was sufficiently low to make it implausible. The NuScale plant has both CIVs in a single valve body. There are no break locations between the valves. However, the weld between the valve body and the CNV safe end is equivalent to those to which break exclusion applies. Therefore, NuScale has extended this boundary outside the CNV to include:

The outboard weld at the CIV
The outboard check or excess flow check valve nozzle weld in pressurizer spray, injection, and discharge lines
DHRS piping welds outside the CNV Accordingly, the guidance of BTP 3-4 B.A.(ii) is used in piping design to ensure that breaks and leakage cracks can be excluded in the containment penetration area. BTP 3-3 non-mechanistic breaks of MSS and FWS piping are also addressed. The remaining high energy piping under the bioshield applies 2 3.6-14 Revision 4

break location and size, as applied in the NPM bay and the RXB. The length of piping and number of welds inside the NuScale CNV is limited. For the NuScale design, no primary or secondary piping other than about 160 feet of DHRS piping is within the break exclusion zone outside containment. The design pressure and temperature of MSS, FWS, and DHRS piping in the break exclusion zone is the same as for the RCS. Break exclusion is not applied to any of the piping in the RXB outside of the bioshield. 2.1.2.3 Leakage Cracks Leakage cracks are excluded in containment penetration areas where the criteria of BTP 3-4 B.A.(ii) are satisfied. For areas outside the containment penetration area, per BTP 3-4 Paragraph B.A.(v), leakage cracks are postulated unless specific criteria are met. For Class 2 piping, the acceptance criterion is for the calculated stress to not exceed 0.4 times the sum of stress limits given in Subarticles NC/ND-3635. BTP 3.4 B.C.(iii) specifies postulating leakage cracks with a flow area of one-half of a pipe diameter by one half pipe wall thickness in piping in the vicinity of essential SSCs, regardless of system. 2.1.3 Pipe Breaks in the Reactor Building (outside the Bioshield) Within the NPM, there are a number of essential SSC that require protection and relatively small amounts of piping. Therefore, postulated pipe break locations within the NPM or in close proximity to the NPM (i.e., under the bioshield) are specifically addressed by analysis, as discussed in Section 3.6.1.3. Beyond the NPM, there are fewer SSC that require protection and a large amount of high- and moderate-energy piping (See Table 3.6-1). The SSC that require protection are evaluated for effects of line breaks or are separated within compartments of the RXB from areas that contain piping. In addition, the building structure necessary to support the modules and to maintain the integrity of the pool (i.e., the ultimate heat sink) is evaluated. Piping arrangements in the RXB have not been finalized yet. It is appropriate, therefore, for evaluation of potential rupture locations beyond the reactor pool bay wall, to identify the bounding dynamic effects of postulated breaks and then to determine if protection is required. The approach is to evaluate:

blast, unconstrained pipe whip, and jet impingement caused by rupture of a main steam pipe.
subcompartment pressurization, spray wetting, flooding, and other adverse environmental effects caused by main steam or CVCS breaks that are potentially limiting where they might occur in the building.

2 3.6-15 Revision 4

A break in a high-energy MSS or FWS line in the RXB (outside of the bioshield) could potentially cause breaks or leakage cracks in smaller diameter or pipe schedule lines of other NPMs, introducing an additional transient in a second NPM. Therefore, RXB MSS and FWS pipes must be arranged, and/or pipe whip restraints must be provided to prevent a collateral rupture, or pipe whip impact analysis must be performed to show that a collateral rupture does not occur. However, the effects of an MSS or FWS break are assumed to cause an MSS bypass line rupture in an adjacent module in order to determine bounding dynamic effects and to ensure that the RXB structure is adequate for beyond design basis interactions between adjoining modules. Once piping arrangements are finalized, the need for pipe whip restraints and barriers may be determined to avoid multi-module effects. This is addressed by the COL applicant as part of COL Item 3.6-3. The CVCS lines in the RXB (outside the bioshield) are not co-located with essential SSC, with the exception of the RXB itself. Therefore, dynamic effects are addressed on a bounding basis and individual break locations are not specified. For flooding and environmental effects, as discussed in Sections 3.4 and 3.11 respectively, breaks are postulated to occur anywhere on the line. Item 3.6-3: A COL applicant that references the NuScale Power Plant design certification will perform the pipe rupture hazards analysis (including dynamic and environmental effects) of the high- and moderate-energy lines outside the reactor pool bay in the Reactor Building (RXB), and update Table 3.6-2 as appropriate. This includes an evaluation and disposition of multi-module impacts in common pipe galleries, and evaluations regarding subcompartment pressurization. The COL applicant will show that the analysis of RXB piping bounds the possible effects of ruptures for the routings of lines outside of the RXB or perform the pipe rupture hazards analysis of the high- and moderate-energy lines outside the buildings. 2.1.4 Pipe Breaks in the Control Building There are no high-energy lines in the CRB. Flooding and environmental evaluations are described in Section 3.4 and 3.11, respectively. 2.1.5 Pipe Breaks in the Radioactive Waste Building There are no high-energy lines or essential equipment in the RWB. Therefore, no breaks or leakage cracks are postulated. 2.1.6 Pipe Breaks Onsite (Outside the Buildings) As discussed in Section 3.6.1.1.6, there are four high-energy lines outside of the RXB and CRB: MSS, FWS, auxiliary boiler, and extraction steam, and multiple moderate-energy lines (See Table 3.6-1). However, there is no essential equipment outside of the RXB or CRB. The routing of piping outside of the RXB, CRB, and RWB is the scope of the COL applicant. Item 3.6-4: Not used. 2 3.6-16 Revision 4

The criteria used to determine the axial locations of postulated pipe breaks are described in Section 3.6.2.1.1, Section 3.6.2.1.2, and Section 3.6.2.1.3. At these locations, either a circumferential or longitudinal break, or both, are postulated according to the following criteria:

For piping sizes larger than NPS 1, at piping terminal ends, a circumferential break only is postulated.
For piping sizes larger than NPS 1 but smaller than NPS 4, at intermediate locations (i.e., not terminal ends), a circumferential break only is postulated.
For piping sizes NPS 4 and larger, at intermediate locations (i.e., not terminal ends), both a circumferential and longitudinal break are postulated unless the location of the break is selected using stress analysis per the criteria given in Section 3.6.2.1.1, Section 3.6.2.1.2, and Section 3.6.2.1.3 and a further evaluation of the stress results is used to determine the break type as follows:

If the circumferential stress range is at least 1.5 times the axial stress range, only a longitudinal break need be postulated If the axial stress range is at least 1.5 times the circumferential stress range, only a circumferential break need be postulated Where circumferential breaks are postulated, the following assumptions are made:

A circumferential break results in pipe severance and separation amounting to at least a one-diameter, lateral displacement of the ruptured piping sections unless physically limited by piping restraints, structural members, or piping stiffness as may be demonstrated by inelastic limit analysis (i.e., a plastic hinge not developed in the piping).
Pipe movement is initiated in the direction of the jet reaction and whipping occurs in a plane defined by the piping geometry and configuration.

Where longitudinal breaks are postulated, the following assumptions are made:

A longitudinal break results in an axial split without pipe severance. Splits are postulated to be oriented (but not concurrently) at two diametrically opposed circumferential locations such that the jet reactions cause out-of-plane bending of the piping configuration. Alternatively, a single split is assumed at the location of highest tensile stress as calculated by detailed stress analysis (e.g., finite element analysis).
Pipe movement occurs in the direction of the jet reaction unless limited by piping restraints, structural members, or piping stiffness as may be demonstrated by inelastic limit analysis.

Longitudinal cracks are not applicable in the CNV, because piping NPS 4 and larger meets LBB criteria. Also, longitudinal breaks are not considered under the bioshield, based on meeting criteria for not considering circumferential breaks. In the rest of the RXB, effects of longitudinal breaks (with break flow areas equal to the piping flow area) are bounded by circumferential breaks. 2 3.6-17 Revision 4

For high-energy lines, with the exception of those portions of piping exempted using the criteria contained in BTP 3-4 B.A(ii) as described in Section 3.6.2.1.2 and Section 3.6.2.7, leakage cracks are postulated at locations that result in the most severe environmental consequences unless otherwise selected by stress analysis. For lines where stress analysis has been performed, postulated leakage crack locations are determined according to the criteria in BTP 3-4 B.A(v) as follows:

For ASME Code, Section III, Class 1 piping at axial locations where the calculated stress range by Eq. (10) in NB-3653 exceeds 1.2Sm.
For ASME Code, Section III, Class 2 and 3 piping, or nonsafety-class (i.e.,

non-ASME Class 1, 2, or 3), at axial locations where the calculated stress equal to the sum of Eq. (9) and Eq. (10) in NC/ND-3653 exceeds 0.4 times the sum of the stress limits given in NC/ND-3653. For moderate-energy lines, leakage cracks are not postulated inside the containment or outside the containment under the bioshield. Per BTP 3-4 Part B.B(iv), leakage cracks need not be postulated in moderate-energy piping located in an area in which a break in high-energy piping is postulated, provided such leakage cracks would not result in more limiting environmental conditions than the high-energy piping break. For the areas inside containment (described in Section 3.6.1.1.1) and outside containment under the bioshield (described in Section 3.6.1.1.2), the effects of leakage cracks in the moderate-energy RCCWS, CFDS, and CES, are bounded by breaks in high-energy lines. In other areas of the plant, ruptures of moderate-energy lines are assumed at locations that result in the most severe environmental consequences. Environmental conditions are based upon the leakage cracks of the worst case (typically largest or hottest) line in the proximity of safety-related SSC. For flooding analysis, full circumferential breaks in piping larger than NPS 2 in a room where they are located are used to evaluate flooding. Environmental effects are discussed in Section 3.11 and flooding analysis is described in Section 3.4. Per BTP 3-4 C(iii)(1) leakage cracks in high- and moderate-energy lines need not be postulated in NPS 1 and smaller piping. Where leakage cracks are postulated in high- and moderate-energy lines, the following criteria from BTP 3-4 C(iii) are applied or are shown to be bounded:

For high-energy piping, the leakage cracks should be postulated to be in the circumferential locations that result in the most severe environmental consequences. For moderate-energy piping, leakage cracks should be postulated at axial and circumferential locations that result in the most severe environmental consequences (per BTP 3-4 B(iii)(2)).
Fluid flow from a leakage crack should be based on a circular opening of area equal to that of a rectangle one-half pipe diameter in length and one-half pipe wall thickness in width. The flow from a leakage crack should be assumed to result in an environment that wets the unprotected components within the compartment with consequent flooding in the compartment and communicating compartments. Flooding effects should be determined on the 2 3.6-18 Revision 4

2.2 Effects of High- Energy Line Breaks In accordance with SRP Section 3.6.2, the dynamic effects of postulated high-energy line break are evaluated using the methodology as described in this section. 2.2.1 Blast Effects The potential for a blast wave to occur depends on the surrounding environment. Key factors include the timing of the break and the initial system thermodynamic conditions. The timing of opening of the break and the initial, intact system thermodynamic conditions also are key factors. Although pipe rupture times of less than a millisecond are unlikely, break opening time is assumed to be instantaneous, maximizing blast formation. The formation and effects of a blast wave caused by an HELB is evaluated using three-dimensional computational fluid dynamics (CFD) modeling that reflects the postulated break characteristics and NuScale plant geometry. The analysis is performed using ANSYS CFX. The acceptability of using CFX for this purpose was demonstrated by performing verification and validation using eight test problems that exercised different capabilities of the code. Key observations from this blast wave modeling are:

A blast wave is weakly formed if the surrounding environment is at low pressure (less than 1 psia), as is the case inside the CNV. Buildup of pressure as blowdown progresses is not relevant, because the blast wave is a prompt and short-lived phenomenon.
The severity of a blast depends on the amount of fluid that can escape within approximately one millisecond of break onset because the blast wave forms within that time.
The NuScale high-energy, steam-filled lines are relatively small, which limits the severity of the blast pressure. The energy available to form the blast is less than one-twenty-fifth that for a typical large, light-water reactor.
Blast waves are not significant for subcooled discharge, because liquid flashing occurs on time scales exceeding that of blast wave formation (Reference 3.6-22).
A blast wave has well-defined and interrelated characteristics. For example, its peak pressure and speed decrease with distance from its origin.
The pressure load applied by a blast wave is of short duration (i.e., an impulse load) and does not apply uniformly across large SSC at a given instant.

Therefore, assuming the peak blast pressure is applied across the entire projected area of a component is inappropriate. The CFD analysis explicitly accounts for the time-varying pressure of the rapidly propagating blast wave. 2 3.6-19 Revision 4

than a flat surface perpendicular to a line between the blast origin and surface. The pressures applied to surfaces by reflection can exceed the incoming wave pressure. For this reason, use of representative plant geometry is necessary. The CFD analysis includes the interaction of incident and reflected waves with each other and nearby surfaces, including how the shape and orientation of surfaces affect reflection.

A small target has a lower peak pressure due to clearing, which is a phenomenon where some of the blast overpressure is relieved by bleeding off around the edge of the target. Because of both pressure-relieving clearing and the short load duration as a supersonic blast wave moves over them, small structures are not exposed to significant loading. The only SSC in the CNV or RXB that are large are the structures (e.g., CNV, RPV, and RXB walls and floors).

The CFD analysis considers clearing.

Several locations and directions of CVCS breaks in the CNV and MSS breaks in RXB pipe gallery were modeled. These were selected to maximize blast pressure on nearby SSC (e.g., close to walls and corners) in order to bound final piping arrangements.

Blast analyses results show a maximum total force of 6000 lbf on a component in the CNV. The maximum total force is less than 10,000 lbf on a component and about 100,000 lbf on the five-foot-thick pool wall in the RXB pipe gallery (the wall load is spread over an area with a radius of about 100 inches, corresponding to an average pressure of less than 14 psig, compared to a concrete compressive strength of 5000 psia). These forces are impulse loads that last only a few milliseconds or less. In summary, three-dimensional CFD analysis of blast wave formation in the CNV and RXB is performed using modeling assumptions that bound the pressurization effects that occur for HELB in the plant. Blast wave force time histories are calculated for nearby SSC. The results show:

Peak forces are low and bounded by the jet thrust forces that subsequently develop. The values are low because NuScale HELB are relatively small diameter and deposit only a small amount of mass and energy in the time it takes for a blast wave to form. The forces inside the CNV are low because the initial low ambient pressure does not support formation of a significant blast wave.
The forces of the passing shock wave are of very short duration.

Therefore, effects of HELB-induced blast waves in the NuScale plant are considered negligible. No damage to surrounding SSC occurs because these loads are small and brief. 2.2.2 Pipe Whip The methodology for pipe whip includes determination of whether a pipe has sufficient energy to whip, whether a whipping pipe can actually contact a 2 3.6-20 Revision 4

of an impact should the previous steps not obviate the possibility of damage. The thrust force caused by release of fluid from a circumferential break of a high-energy piping system may cause the piping to rotate about a plastic hinge-point (e.g., pipe restraint, pipe anchor point) and possibly impact nearby SSC. Inside the CNV, the largest pipe size subject to HELB conditions is NPS 2 and target SSC are robust [e.g., reactor vent valves (RVVs)]. High-energy piping systems larger than NPS 2 have been qualified for LBB inside the CNV. Outside the CNV, under the bioshield, piping satisfies the criteria of BTP 3-4 B.A.(ii) or (iii) to conclude that no breaks occur and that piping does not need to be evaluated for whip. However, nonmechanistic breaks of MSS and FWS lines and leakage cracks are considered. In the RXB outside the bioshield, MSS, FWS, and CVCS lines are subject to a postulated HELB, but there are only a limited number of SSC requiring protection. Also, Auxiliary Boiler System (ABS) line leakage cracks are evaluated. The following considerations apply to evaluation of pipe whip:

For piping meeting the criteria of break exclusion or LBB, pipe whip is not considered because dynamic effects of ruptures are excluded.
If the end is an RPV or CNV safe end, whip does not occur because the safe end and its nozzle is short, stiff, straight, and restrained by the component.
In accordance with SRP Section 3.6.2, a pipe struck by another pipe of equal or smaller diameter and schedule (i.e., wall thickness) does not break or crack. In the CNV where HELB are limited to NPS 2 Schedule 160 pipe, the RPV, CNV, ECCS valve bodies, and CRDMs are each equivalent to larger, thicker-walled pipe and, therefore, do not crack or break.
Where pipe ruptures are postulated to occur, the distance is determined from the break location to the nearest restraint that limits the range and/or direction of the pipe whip.
The jet thrust necessary to cause pipe whip is determined. The calculation of thrust and jet impingement forces considers no line restrictions (e.g., a flow limiter) between the pressure source and break location, but does consider the absence of energy reservoirs, as applicable (e.g., the high-point vent pipe in the CNV is normally isolated).
If the jet thrust is insufficient to yield the pipe, then pipe whip at that break location is eliminated from further consideration.
Pipe whip is considered to result in unrestrained motion of the pipe along a path governed by the hinge mechanism and the direction of the vector thrust of the break force. A maximum rotation of 90-degrees is assumed about a hinge. Pipe whip occurs in the plane defined by the piping geometry and configuration and initiates pipe movement in the direction of the jet reaction, as identified in BTP 3-3.

2 3.6-21 Revision 4

inside the CNV. The minimum wall thickness of these components is at least three times that of the postulated whipping pipe. In view of the SRP 3.6.2 provision for impact of a pipe on like-size or larger pipe, the RPV, CNV, CRDMs, and ECCS valve bodies neither rupture nor crack if struck by a whipping NPS 2 Schedule 160 pipe in the CNV. Because of the disparity in the thickness of the walls, the whipping pipe kinetic energy is absorbed in the bending and crushing of the pipe itself. Functionality of components with moving parts (i.e., CRDMs and ECCS valves) following impact is addressed. Postulated break locations are at the RPV (head for spray and high-point vent degasification lines, and side wall for injection and discharge lines) and CNV heads. The high-point vent line does not whip for a break at the RPV head, because the isolated line is filled with steam that immediately depressurizes as the break begins to open. Ruptures above the NPM under the bioshield are excluded and there are no safety-related or essential components with whip range elsewhere in the RXB. However, the RXB structural integrity and, in particular, the integrity of the pool wall must be assessed, so pipe whip impact force on concrete surfaces is determined. After break-opening, the steady-state jet thrust force, Fb, is: Fb = CTPoAe Eq. 3.6-1 where, Fb = Steady state thrust force at the break (lbf), CT = Thrust coefficient (unitless), Po = Internal system pressure (psia), and Ae = Pipe flow area (in2).

Applying the jet thrust force at the distance to the nearest restraint or anchor point determines the force available to overcome the pipe plastic bending moment and accelerate the pipe. If there is sufficient energy for whip to occur, the energy to yield the pipe is not deducted from its kinetic energy. Pipe whip evaluations have determined that impact on targets in the CNV is unlikely.

Quantitative pipe whip impact evaluation is performed only for concrete walls and other structures in the RXB.

To determine the depth of penetration of the whipping pipe into an RXB wall, the Sandia formula developed by Young (Reference 3.6-20) is used. For an assumed pipe whip segment length and angle of travel that is bounding, whip of an MSS line (the highest-energy pipe in the RXB) causes a penetration of a depth of 4.1 inches, or about 7 percent of the minimum wall thickness of 2 3.6-22 Revision 4

smaller pipe size, chemical and volume control system pipe impacts are even less damaging. 2.2.3 Jet Impingement Target SSC in the path of jets issuing from postulated breaks are assessed for the load imparted by the jet. In industry testing, single-phase steam jets with upstream pressures of 1200 psia were found to cause damage to pipe insulation at a distance of up to 25 times the pipe exit diameter (i.e., L/D = 25). However, insulation is fragile as evident from Reference 3.6-16, which reports types of insulation suffering damage due to impingement pressures as low as 4 psig. NUREG/CR-6808 (Reference 3.6-17) Table 3-1 provides the impingement pressures found in testing to cause damage to various types of piping insulation used in U.S. pressurized water reactors. The damage pressures range from 4 to 40 psi for fibrous insulation to a high of 190 psi for two types of reflective metal insulation. Insulation is more fragile than the uninsulated solid metal surfaces of SSC inside the CNV. Therefore, jet impingement pressures need to be considerably above 190 psi to be of concern. Impingement loads are only relevant for hard or relatively hard targets such as ECCS valve bodies, the CNV steel wall, and the RXB concrete structure. Impingement pressures must be substantial (above 190 psia) rather than the less than 4 psia needed to protect against dislodging insulation. As such, fewer uncertainties exist in predicting jet impingement effects on piping, and the relevant penetration distance is much shorter than 25 L/D. Jet impingement testing was performed on electrical cable in support of the AP1000 assessment of debris generation. The conclusion was that cables at greater than or equal to 4 L/D from a jet simulating an AP1000 LOCA were not damaged. The results were given in terms of distance because of difficulty in accurately measuring impingement pressure. The NRC staff agreed with the conclusion. In Reference 3.6-18, the Advisory Committee on Reactor Safeguards (ACRS) also agreed, stating, The recommended distance of four break diameters from a loss-of-coolant accident jet, at which unprotected cables would not be damaged, has been shown by testing to be sufficiently conservative to bound plant conditions with high likelihood. Although the focus of this testing did not include cable functionality, inspection of test target cables showed no damage at greater than or equal to 4 L/D (with exception of one cable). The results were applicable only to the type of cables actually tested, but an AP1000 LOCA jet is considerably larger and higher energy than a NuScale NPS 2 HELB. Therefore, it is likely that even unprotected cable inside the CNV would survive jet impingement from an NPS 2 HELB provided its separation from the break exit exceeded 4 L/D, or 6.75 inches. NuScale cable to be used in the CNV is tested for survival under jet impingement. 2 3.6-23 Revision 4

maximum force of the jet and its maximum pressure is that at the break exit, or 103,000 lbf and 630 psia, which is well within the minimum 5000 psi compressive strength of the concrete making up the five-foot thick wall. In addition, the effect of erosion is negligible. An overview of the NuScale resistance to jet impingement is:

The damage potential of the smaller-scale NuScale piping is reduced compared to large reactors:

Based on plant operating conditions and size of piping, thrust loads for NuScale line breaks are a fraction of those encountered in large LWRs (e.g., a NuScale 12-inch MSS line has about five percent of the total thrust force of a 38-inch MSS line break). Main steam system HELB occurrence is limited to the RXB, because MSS breaks inside the CNV and under the bioshield are eliminated by LBB and break exclusion, respectively. Considering MSS steam density, flow rate driven by the system to ambient differential pressure, and the full break single-ended flow area, the NuScale MSS HELB mass and energy transfer is approximately five percent of that in other large LWRs.

Jet reaction load and, if within the ZOI, potential jet impingement load is included in load combinations in accordance with FSAR Section 3.9 and Section 3.12.
Damage to insulation on piping is not a concern:

In the CNV, no pipe or component thermal insulation is used. Under the bioshield, no ruptures are postulated. In the NPM outside the pool area, dislodged insulation has no effect on long-term NPM cooling. Thus, allowable impingement pressure on SSC is considerably higher than that in large pressurized water reactors where insulation stripping is relevant.

The maximum load imposed by the impinging jet is that of the thrust force of the broken pipe at the break exit.

Because only NPS 2 RCS pipes are locations of postulated breaks in the CNV, the load is limited to the maximum operating pressure times the flow area times the thrust coefficient (1.26 for steam and two-phase jets). The total load imposed by the jet is approximately 5220 lbf. The applied load is adjusted by a target shape factor (e.g., 0.576 for a jet striking a cylinder normal to its axis) and by the cosine of the angle from perpendicular for the intersection of the jet and the target surface. These two adjustments reduce the imposed load to below 2000 lbf, or approximately two times the weight of a reactor recirculation valve. Finally, the jet rapidly traverses the zone of influence (ZOI) caused by whip of the broken pipe, moving more than 100 ft/sec within a few degrees of 2 3.6-24 Revision 4

only transiently. The RVVs are approximately a foot in diameter, meaning that they are within the jet for a maximum of 0.01 of a second. Exposing a 1000 lbm, thick-walled, metal component to 2000 lbf for 0.01 of a second or less is a negligible load that can be omitted from load combinations that include dead weight and seismic accelerations of over 10 g.

The impingement damage threshold of 190 psi is a sufficient measure of the structural integrity of components, but does not confirm functionality.

Essential components inside the CNV are qualified for a CNV design condition of at least 1000 psia saturated steam. This exceeds the 190 psi impingement acceptance threshold of 190 psia by a factor of more than five and is sufficient basis to consider functionality after jet impingement to be demonstrated.

Jet impingement on concrete is neither a pressure load nor an erosion concern.

Having addressed the resistance of the NuScale design to jet impingement damage, the HELB jet conditions must be determined. Three categories of jets are considered:

1) Liquid jets

2) Two-phase jets

3) Steam jets As discussed for other effects, jet behavior and effects differ for the three areas of the plant:

Inside the CNV: breaks are limited to NPS 2 RCS-connected and DHRS piping because the SGS piping meets LBB. Only a degasification line break is steam, however, the reverse flow from a pressurizer spray line break almost immediately turns to steam. Other breaks such as DHRS, the injection line, or spray line forward flow are two-phase.
Under the bioshield: piping satisfies criteria that no postulated breaks occur.
In the RXB, outside the bioshield: piping arrangements are not finalized, so break locations and jet directions are assumed to be throughout in the rooms containing high-energy piping. The piping is limited to NPS 12 and 4 MSS, NPS 6 FWS, and NPS 2 to 3 CVCS piping at various pressures and temperatures (see Table 3.6-1 and Table 3.6-4). Main steam system jets are steam only, whereas FWS and CVCS breaks are two-phase.

The concern for jet impingement that underlies regulatory guidance is the stripping of insulation with subsequent sump blockage as described in GSI-191. As noted above, the impingement damage threshold for NuScale is greater than 190 psig. 2 3.6-25 Revision 4

Liquid jets are assumed to not expand (i.e., the cross section of the pipe rupture is maintained) and to not droop with distance (i.e., travel straight until impeded). Additionally, a 2.0 thrust coefficient is used for dynamic loading. The only areas subject to liquid jets are in the RXB where CVCS lower temperature, high-pressure piping is present. The essential SSCs in this area are the CVCS demineralized water makeup valves and RXB structure (liquid jets are considered to have less potential to damage concrete structure than steam jets, which are shown to be acceptable). Two-phase jets Two-phase jets are assessed using the methodology of NUREG/CR-2913 (Reference 3.6-19). A bounding approach is taken by identifying criteria for jet formation in order to avoid the need to analyze individual break locations in the CNV and RXB.

In the CNV Although the low operating pressure of the CNV is a variation from the experimental and analytical basis of NUREG/CR-2913, the low ambient pressure results in faster expansion of the jet and is conservative when estimating loading.

Only RCS-connected NPS 2 pipe breaks are evaluated (DHRS system pressure and temperature are lower at postulated break locations). The inputs needed for the NUREG/CR-2913 methodology are the system static thermodynamic conditions, as shown in Table 3.6-4. Following the methodology, the relevant graph of Appendix A of NUREG/CR-2913 is selected to obtain target pressure and total force on the target for appropriate values of P0, T0, or X0, and distance to the target in L/D. For the CVCS breaks in the CNV, the thermodynamic conditions are 48 degrees K subcooling and 67 bar. The appropriate graph is Figure A.39, which shows pressures at specific points downstream in L/D and radially from the jet centerline in r/D. At the origin of the plot is the jet centerline at the break exit plane, and the shaded area at the lower left is the jet core (the region that has not yet begun to interact with the environment and in which fluid striking a target would experience full recovery of the fluid stagnation pressure). The letters A through D refer to the key for pressure (letters E and beyond for pressures above 10 bar are not plotted because they exist only near the jet core). For example, a letter B indicates pressure is 2.5 bar at 4 L/D and 1.5 r/D. The jet core is the region immediately downstream of a break in which the target pressure is the full stagnation pressure. Reference 3.6-17, Section 3.3.1.1 states that this region is significant only for jets involving subcooled stagnation conditions. Figure A.39 of NUREG/CR-2913 shows that the jet core dissipates within 2 L/D or about 3.4 inches for a thermodynamic condition similar to a chemical volume and control system HELB. This is viewed as conservative. Reference 3.6-13 Section 3.5.3.B notes that Sandia (Reference 2.6.19) 2 3.6-26 Revision 4

the core length will always be longer than 0.5 D for subcooled and saturated water jets. Reference 3.6-13 notes, however, that test data is not consistent with the Sandia model, with only one or two test data sets exhibiting something like a liquid core while most data contradict the presence of a liquid core. Reference 3.6-13 concludes If a liquid core exists, it seems to be much smaller than indicated by Sandia. At 2.5 L/D and 1 r/D, the single D point is a pressure of 10 bar (145 psig), below the NuScale damage threshold of 190 psig. Within 4 L/D or about 6.8 inches, the jet peak pressure has dropped to below 5.0 bar (72.5 psig). The A points representing 1.0 bar correspond to the edge of the jet. The jet persists beyond 7.5 L/D, which is indicative of the concern for fibrous insulation damage at pressures of 4 psig out to a 10 L/D penetration distance. For NuScale's design, pressures at about 2 L/D are low enough to cause no damage to the hard components. Although the NUREG/CR-2913 figure can be used to determine the ZOI, the ZOI in the CNV is assumed to be in the forward-facing hemisphere because of the greater spreading angle in the low-pressure CNV and possible pipe whip.

In the RXB Similarly, for chemical and volume control system HELB in the RXB, the generic approach of a universal ZOI allows for breaks at locations determined once pipe routing is finalized and for pipe whip. Based on the discussion that follows for steam jets, CVCS pressure loading, as shown in Figure A.39 of NUREG/CR-2913, is not damaging.

Steam Jets

In the CNV For breaks inside the CNV, expansion of the jet into the low-pressure surroundings results in different behavior than is experienced for HELB. Wider jet spreading (a half-angle exceeding 60 degrees) is expected to occur, because the initially low air density of the CNV removes most of the resistance to jet expansion. The wider jet expands the ZOI, but substantially reduces the pressure and the penetration length, because the mass and energy of the jet are widely dispersed. Although pressure within the CNV increases with time, the pre-existing wide expansion of the jet persists because the jet is already established.

For simplicity and because there are no rigid restraints at postulated break locations to constrain separation, circumferential breaks are assumed to be full separation. For circumferential breaks with full separation, it is assumed that an essential system or component is within the ZOI if it is located within the forward-facing hemisphere based on the original pipe orientation. 2 3.6-27 Revision 4

pressure is assumed to decrease with distance proportional to the area of a jet that expands at a 30-degree half-angle to five pipe diameters and then at 10 degrees beyond that. A half-angle of 30 degrees is less than identified in the ANSI/ANS 58.2 Standard and in other jet analyses for expansion into surroundings at normal atmospheric pressure. Thus, the jet pressure is below the 190 psi threshold for component damage at 2.2 L/D (3.65 in.). Although the NRC has challenged the general applicability of the ANSI/ANS Standard 58.2 spreading model, a half-angle of approximately 45 degrees or more is usually used. As the jet spreads more rapidly into the low-density CNV atmosphere, a 30-degree assumption is sufficiently conservative to bound actual jet impingement pressures due to local variation within the jet. As noted, the jet core is only significant for subcooled jets. Reference 3.6-19 Section 3.6 discusses the core length Lc as 1/2D, one half of the pipe diameter for saturated stagnation conditions. It also notes that the length Lc depends on the time it takes a pressure wave to travel from the outer edge of the nozzle (i.e., break) to the jet center. Figure 4.3 of Reference 3.6-19 shows that for zero degrees subcooling Lc=1/2D. Thus, even if a jet core existed for a steam jet, its influence would be dissipated within 1/2D, which is too close for a jet impingement force to be of concern compared to pipe whip impact.

In the RXB Jet core length is not relevant for RXB breaks because full exit plane pressure is assumed. The distance between a break and a target SSC is not defined because RXB piping arrangements have not yet been finalized. To verify suitability of the design of the RXB, bounding HELB scenarios have been identified.

The MSS lines are larger and contain more energy than other potential jet sources in the RXB. Demonstrating passing performance for MSS breaks provides confidence that final HELB analysis results are bounded. Therefore, a conservative approach is taken in which the jet impingement pressure is assumed to be the same as that at the break exit (i.e., no reduction for spreading with distance). For a main steam system HELB, the break exit pressure is 500 psia. Applying the thrust coefficient CT of 1.26 yields a jet impingement pressure of 630 psi, or about one-eighth of the minimum compressive strength of the concrete and less than the previously discussed erosion that testing demonstrated is acceptable. Jet impingement for HELB in the NuScale plant is therefore not a source of concern because of the lesser jet energies associated with the smaller size piping, and because of the high impingement pressure damage threshold associated with not needing to protect against insulation being dislodged. 2 3.6-28 Revision 4

Based upon concerns raised by the ACRS in 2004, the NRC identified (SRP Section 3.6.2) that unsteadiness in free jets, especially supersonic jets, tends to propagate in the shear layer (i.e., the region with a large velocity gradient near the boundary of the jet) and induce time-varying oscillatory loads on obstacles in the flow path. The ACRS concern was that pressures and densities vary nonmonotonically with distance along the axis of a typical supersonic jet, feeding and interacting with shear layer unsteadiness. In addition, for a typical supersonic jet, interaction with obstructions could lead to backward-propagating transient shock and expansion waves that cause further unsteadiness in downstream shear layers. The concern was that synchronization of the transient waves with the shear layer vortices emanating from the jet break could lead to amplification of the jet pressures and forces (a form of resonance) that is not considered in ANSI/ANS 58.2. Should the dynamic response of the neighboring structure also synchronize with the jet loading time scales, further amplification of the loading occurs, including at the source of the jet. General observations by investigators were that strong discrete frequency loads occur when the impingement surface is within 10 diameters of the jet opening, and that when resonance within the jet does occur, amplification of impingement loads might result. The basis for this concern was research into such amplification of loads that occur in the interaction of the jet issuing from vertical and short take-off and landing aircraft and certain industrial gas jet applications. It causes vibration and fatigue damage to aircraft parts, jet deflectors, parts cleaned with gas jets, etc. This phenomenon has been studied extensively, with considerable work performed to mitigate its effects. Experiments simulating HELB routinely evince random oscillations, but not resonance. For dynamic amplification and resonance to occur, a number of criteria must be met. These criteria are based on the research referenced in SRP Section 3.6.2 and similar work that identified the physical phenomena leading to resonance. These processes require a stable, axisymmetric jet impinging at a fixed distance perpendicular to a large, flat surface. The processes at work during a HELB have fundamental differences from those that occur in a jet with dry, noncondensable gas issuing from a smooth, fixed nozzle. These physical differences involve instability of the discharge, irregular discharge geometry, phase changes that suppress pressure changes, misalignment of jet and impingement target surface preventing establishment of a feedback loop, lack of an appropriately flat surface within a sufficiently close distance, and etc. If one of these criteria is not met, a resonance is implausible. In a HELB in the NuScale plant, none of the criteria is satisfied, precluding the formation of a resonance. Specifically, each of the following characteristics of postulated HELB in a NuScale plant is sufficient to ensure a resonance does not occur.

A whipping pipe either 1) comes to rest against an object that intercepts a portion of the jet and distorts its axisymmetry or 2) flutters, causing a variation 2 3.6-29 Revision 4
The break exit is distorted because of tearing as the break opens, which eliminates axisymmetry.
Jets in the CNV dissipate in a short distance. The plant geometry precludes the end of a break coming within 2 L/D of a suitable impingement surface. Beyond that distance, the jet has weakened too much for amplification to be a problem, even if it does occur.
No suitable (i.e., even, flat) impingement surfaces exist within the CNV.

Relevant SSC are curved, which redirects reflected acoustic energy away from the break exit.

The presence of a steam/water mixture in the jet acts to dampen pressure oscillations, preventing amplification.
Splashing from the jet (and the jet from the opposite end of the break) interferes with the stability of the jet.

2.2.5 Subcompartment Pressurization In the CNV, pressurization from postulated HELB is bounded by ECCS initiation and no breaks for which dynamic effects must be considered are postulated under the bioshield. For the RXB, bounding HELB scenarios have been identified based on the high-energy systems in the subject areas of the building. The largest mass and energy input considered is in the pipe gallery and involves a MSS HELB with pipe whip that causes a MSS NPS 4 bypass rupture. For each scenario, the necessary vent path area to avoid high subcompartment pressure is identified and verified to be provided by the RXB design. The allowable differential pressure across building structural elements (e.g., walls) is set to ensure that building and reactor pool structural integrity is satisfactory. 2.3 Protection Methods As discussed previously, methods employed in the NuScale design to address pipe ruptures vary by location and system.

In the CNV, main steam and feedwater piping is designed to satisfy LBB. Reactor coolant system-connected intermediate piping locations are designed to satisfy criteria to avoid breaks, while terminal ends of RCS lines are analyzed for break effects.
Above the NPM under the bioshield, breaks are excluded by identifying a design that satisfies criteria for break exclusion in the containment penetration areas or criteria to avoid breaks at the intermediate piping locations.
In the RXB, the SSC requiring protection against rupture effects are generally separated in rooms not containing high- or moderate-energy piping, and bounding analysis is performed to ensure the structural integrity of the RXB itself.

2 3.6-30 Revision 4

smaller-scale systems reduce the amount of energy available to drive blasts, pipe whip, and jet impingement. Short piping lengths, intervening obstacles, short jet reach, and hard targets resistant to damage lower the risk for interactions that could adversely affect the functionality of safety-related and essential SSC. 2.3.1 Restraints, Barriers, and Shields Pipe whip restraints may be used to limit the motion of a broken pipe to prevent it from hitting an essential structure, system, or component. Protection for pipe whip and jet impingement is also available through barriers afforded by walls, floors, and other structures. Sufficiently large and robust SSC can also function as a pipe whip barrier or jet impingement shield. 2.3.1.1 Pipe Whip Restraints Pipe whip restraints constrain movement of a broken pipe for purposes of preventing or limiting the severity of contact with essential SSC. Restraints installed only for purposes of controlled pipe whip are not ASME Code components; restraints that also serve a support function under normal or seismic conditions are designed to ASME criteria. The design criteria for pipe whip restraints are:

Pipe whip restraints do not adversely affect structural margin of piping for other conditions.

Restraint design does not restrict thermal expansion and contraction. The restraint design either: a) does not carry loads during normal operation or seismic events or b) the structural analysis includes a conservative load combination.

Pipe whip restraints are located as close to the axis of the reaction thrust force as practicable. Pipe whip restraints are generally located so that a plastic hinge does not form in the pipe. If, due to physical limitations, pipe whip restraints are located so that a plastic hinge can form, the consequences of the whipping pipe and the jet impingement effect are further investigated. Lateral guides are provided where necessary to predict and control pipe motion. For further details, see the Pipe Rupture Hazards Analysis technical report TR-0818-61384.
Generally, pipe whip restraints are designed and located with sufficient clearances between the pipe and the restraint, such that they do not interact and cause additional piping stresses. A design hot position gap is provided that allows maximum predicted thermal, seismic, and seismic anchor movement displacements to occur without interaction.

Exception to this general criterion may occur when a pipe support and restraint are incorporated into the same structural steel frame, or when a zero design gap is required. In these cases, the pipe whip restraint is included in the piping analysis and designed to the requirements of pipe support structures for all loads except pipe break, and designed to 2 3.6-31 Revision 4

In general, the pipe whip restraints do not prevent access required to conduct inservice inspection examination of piping welds. When the location of the restraint makes the piping welds inaccessible for inservice inspection, a portion of the restraint is designed to be removable to provide accessibility.
Analysis of pipe whip restraints Is either dynamic or conservative static.

Static analysis includes

dynamic load factor of 2.0 to account for the initial pulse thrust force, unless a lower value is analytically justified
potential increase by a factor of 1.1 in loading due to rebound.

Loading combination includes dead weight, seismic, and the jet thrust reaction force The criteria for analysis and design of pipe whip restraints for postulated pipe break effects are consistent with the guidelines in ANSI/ANS 58.2-1988. Design is based on energy absorption principles by considering the elastic-plastic, strain-hardening behavior of the materials used. Non-energy absorbing portions of the pipe whip restraints are designed to the requirements of AISC N690 Code. Except in cases where calculations are performed to determine if a plastic hinge is formed, the energy absorbed by the ruptured pipe is assumed to be zero. That is, the thrust force developed goes directly into moving the broken pipe and is not reduced by the force required to bend the pipe. In that a HELB is an accident (i.e., infrequent) event, pipe whip restraints are single use: allowed to deform provided the whipping pipe is restrained throughout the blowdown. Where structural members of a restraint are designed for elastic response, a dynamic increase factor is used. Allowable strain in a pipe whip restraint is dependent on the type of restraint.

Stainless steel U-bar - this one-dimensional restraint consists of one or more U-shaped, upset-threaded rods or strips of stainless steel looped around the pipe but not in contact with the pipe. This allows unimpeded pipe motion during seismic and thermal movement of the pipe. At rupture, the pipe moves against the U-bars, absorbing the kinetic energy of pipe motion by yielding plastically.
Structural steel - this two-dimensional restraint is a stainless steel frame encircling the pipe that does not restrict pipe motion for 2 3.6-32 Revision 4

energy of the pipe by deforming plastically.

Crushable material - if used, the allowable energy absorption of the material is 80 percent of its capacity based on dynamic testing performed at equivalent temperatures and at loading rates of

                     +/-50 percent of that determined by analysis.

Note that a wall penetration may also serve as a two-dimensional pipe whip restraint, provided the wall has sufficient strength to resist the pipe load.

Material properties are consistent with applicable code values, with strain-rate stress limits 10 percent above code or specification values, consistent with NRC guidance (SRP 3.6.2, III.2.A).

2.3.1.2 Pipe Whip Barriers Standard Review Plan 3.6.2 identifies that an unrestrained, whipping pipe need not be assumed to cause ruptures or through-wall cracks in pipes of equal or larger NPS with equal or greater wall thickness. By extrapolation, a structure, system, or component made of metal of equivalent or better yield strength, equal or larger diameter, and equal or greater wall thickness does not only not leak or crack but also obstructs further travel of the whipping pipe, protecting SSC farther away from being struck. The pipe whip load must be considered for inclusion in SSC load combinations to verify that the barrier is not displaced by pipe whip impact. For any structures added to serve as a barrier (or jet impingement shield), Seismic Category 1 loading is analyzed to confirm the structure does not fail and cause damage. 2.3.1.3 Jet Impingement Shields NRC guidance does not have specific criteria for judging suitability of an SSC as a jet shield. Regarding impingement effects, if the following criteria are met, then the SSC is judged capable of serving as a shield without further evaluation:

The diameter and wall thickness of the shield meet the criteria for a pipe whip barrier with a size equal or greater than that of the broken pipe.
The barrier is of sufficient area and positioned to subtend a solid angle from the pipe break opening (considering potential pipe whip) that covers the essential SSC to be protected.
The barrier is solid (without openings) to the extent that no direct line of sight exists from the break opening to the essential SSC. This criterion allows for some indirect passage of spray through an opening, but environmental qualification for pressurization and flooding demonstrates 2 3.6-33 Revision 4

2.4 Guard Pipe Assembly Design Criteria for Piping in Containment Penetration Areas Guard pipes are not used in the containment penetration area. 2.5 Analytical Methods to Define Forcing Functions and Response Models See Section 3.6.2.2. 2.6 Dynamic Analysis Methods to Verify Integrity and Operability See Section 3.6.2.2. 2.7 Implementation of Criteria Dealing with Special Features See Section 3.6.2.1.2. Connection of Reactor Vent Valves and Reactor Recirculation Valves to the Reactor Vessel In the NuScale design, each of three RVVs and two RRVs bolt directly to the reactor vessel. These five bolted-flange connections are classified as break exclusion areas. Because this configuration does not include a physical piping length, a majority of the BTP 3-4 B.A (ii) criteria do not apply. However, these BTP 3-4 B.A (ii) criteria generically involve design stress and fatigue limits and in-service inspection (ISI) guidelines, which are addressed for these bolted connections below. Additionally, discussion is provided regarding threaded fastener design and leakage detection, to demonstrate that the probability of gross rupture is extremely low. The leakage detection systems along with in-service inspections provide assurance that potential failure mechanisms are detected before the onset of a catastrophic failure involving the fasteners of the bolted flange connections for the RRVs and RVVs, and therefore, that a break at this location need not be postulated. Design Stress and Fatigue Limits BTP 3-4 B.A(ii)(1) specifies more conservative stress and fatigue limits for ASME Class 1 piping in containment penetration areas than those required for piping by ASME Code, Section III, NB-3653. The bases for these more conservative limits include a desire to limit the stresses resulting from service loads (excluding those due to peak stresses) to within the material yield strength (i.e., elastic strains), and a concern that the cumulative usage factor calculation account for the possibility of a faulty design or improperly controlled fabrication, installation errors, and unexpected modes of operation, vibration, and other structural degradation mechanisms. The RVV and RRV bolted connections are not classified as piping by their design specifications, and instead are classified as components designed to the rules of NB-3200. For the RVV and RRV bolt material (SB-637 UNS N07718), the design criteria 2 3.6-34 Revision 4

considering the more restrictive limits of BTP 3-4 B.A(ii)(1). Therefore, the imposition of more conservative stress limits are not justified. Additional limits on CUF are not justified because the risk of a faulty design and fabrication and installation errors for a flanged connection is low compared to that of a piping system. The possible degradation mechanisms applicable to Class 1 piping systems do not apply to the ECCS valve bolts. These considerations are addressed further below. Faulty design is not a concern for the RVV and RRV flanges as the design features for these flanged connections that affect the stresses in the bolts are primarily the number and size of the bolts used, which are selected based on industry standards (ASME B16.5). The RVV and RRV flanged connections consist of Class 2500 NPS 5 and NPS 2 B16.5 flange configurations, respectively. ASME B16.5, "Pipe Flanges and Flanged Fittings," has a history of reliability. In addition to conforming to an industry standard design, detailed analysis is required to validate the design per ASME BPVC Section III, NB-3230, including a fatigue evaluation. The fatigue evaluation for these bolts utilizes the fatigue curve from ASME Section III, Division I, Mandatory Appendix I, Figure I-9.7. Figure I-9.7 was generated specifically for small diameter bolting made of SB-637 UNS N07718. Also, as required by NB-3230.3(c) for high strength bolting, a fatigue strength reduction factor of no less than 4.0 is applied to the bolts. The fatigue strength reduction factor specified for bolting further reduces the risk of a faulty design for the RVV and RRV bolting, as compared to ASME Class 1 piping systems. To address fabrication concerns, additional surface and UT examinations, beyond the ASME code requirements for these components, have been specified to properly control fabrication. Bolts analyzed using NB-3232.3(b) have further requirements as stated in NB-3232.3(b)(2) and (3) that place controls on fabrication, by specifying both a minimum thread root radius and minimum radius between the head and shank, thus ensuring that the specified fatigue strength reduction factor used in the calculation of CUF is sufficiently conservative. Unexpected modes of operation for piping systems in the nuclear industry generally involve thermal stratification, cycling, and striping. These situations do not apply to these valves. Unexpected vibration is another common concern, however, the RVVs and RRVs are within the scope of the NuScale Comprehensive Vibration Assessment Program (CVAP). As described in TR-0716-50439, "NuScale Comprehensive Vibration Assessment Program Technical Report," the CVAP ensures that the structural components of the NPM exposed to fluid flow are precluded from the detrimental effects of flow induced vibration (FIV). Other degradation mechanisms that have contributed to past piping failures and not already discussed are addressed below. Included is an explanation as to why these mechanisms are less likely to occur in the RVV and RRV valves than in a typical piping system.

Corrosion - Not applicable as suitable materials have been selected and the bolts are not exposed to fluid.

2 3.6-35 Revision 4

Stress Corrosion Cracking (SCC) - Not applicable as suitable materials have been selected and the bolts themselves are not exposed to fluid.
Water Hammer - Water hammer is not credible because there is no downstream piping and the valves discharge into a vacuum. Additionally, functional testing is performed for these valves including the dynamic effects of blowdown. Blowdown is classified as a service level B load in the ASME loading combinations for the valves, and therefore is included in the fatigue evaluations of the bolts.

In-Service Inspection BTP 3-4 B.A(ii)(1) states that a 100% volumetric in-service examination of all pipe welds should be conducted during each inspection interval as defined in ASME Code, Section XI, IWA-2400. This requirement is addressed for the RVV and RRV bolting by providing augmented ISI requirements for these bolts that exceed the Code requirements. For in-service inspection, if the connection is disassembled during the interval, a UT inspection is performed on the bolts (Section 3.13.2). If the connection is not disassembled during the inspection interval, a volumetric inspection of the connection is performed in-place. Additionally, exceptions in the ASME code for flanged connections that allow only a sample of bolting to be inspected are not followed, and instead all flange bolts for all RVVs and RRVs are inspected during each inspection interval. Threaded Fastener Design The applicable guidelines and recommendations in NUREG-1339 have been adopted by NuScale. Lubricants containing molybdenum sulfide are prohibited for pressure-retaining bolted joints including the RVV and RRV joints. Of the degradation mechanisms listed in NUREG-1339, only SCC could potentially affect RVV and RRV bolted joints. Alloy 718 is highly resistant to SCC in borated water. To further improve Alloy 718 SCC resistance, the solution treatment temperature range prior to precipitation hardening treatment is restricted to 1800°F to 1850°F. Additionally, the RRV bolting is submerged in borated water only during refueling, at a much lower temperature than RCS operating temperature, further reducing SCC susceptibility. The RVV bolting materials are not submerged in borated water as part of any normal operating condition. Based on these considerations, SCC is unlikely for Alloy 718 studs for RVVs and RRVs. Threaded fastener design is discussed further in DCD Section 3.13. Leakage Detection FSAR Section 3.6.3 and FSAR Section 5.2.5 describe how the reactor coolant pressure boundary leakage detection systems conform to the sensitivity and response time recommended in Regulatory Guide 1.45, Revision 1. Leakage monitoring is provided by two means, the change in pressure within the CNV and collected condensate from the containment evacuation system. Even under a scenario where leakage occurs due to one or more postulated bolt breaks, containment leakage monitoring systems are sensitive to a leak rate as low as 0.01 lbm per minute (or ~0.001 gallon per minute). This is because the containment is a relatively small closed volume and is maintained at a 2 3.6-36 Revision 4

surface (lower surface roughness compared to the crack morphology of fatigue cracks), and a straighter flow path that causes less pressure loss through the flow path in the Henry-Fauske's flow model. Therefore it is expected to result in a higher leak rate than through other postulated LBB fatigue cracks, when other conditions are similar. High containment pressure is also a safety actuation signal that initiates a reactor trip. 3 Leak-Before-Break Evaluation Procedures General Design Criterion 4 includes a provision that the dynamic effects associated with postulated pipe ruptures may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping. This analysis is called LBB. The LBB concept is based on the plant's ability to detect a leak in the piping components well before the onset of unstable crack growth. For the NuScale Power Plant, the application of LBB is limited to the ASME Class 2 main steam and feedwater piping systems inside the CNV. The FWS piping analysis addresses significant feedwater cyclic transients and produces bounding loads for the ASME Class 2 piping with respect to LBB. The methods and criteria to evaluate LBB are consistent with the guidance in Standard Review Plan 3.6.3 and NUREG-1061, Volume 3. Potential degradation mechanisms are described in Section 3.6.3.1; analysis for main steam and feedwater piping is provided in Section 3.6.3.4. Leak detection is discussed in Section 3.6.3.5. 3.1 Potential Degradation Mechanisms for Piping In high-energy piping systems, environmental and operating material degradation could adversely affect the integrity of the system as well as the piping system LBB applicability. The application of LBB requires that the affected systems not be susceptible to environmental and operating degradation mechanisms such as erosion/corrosion, fatigue loads, stress corrosion cracking, creep damage, erosion damage, irradiation embrittlement or water hammer. These mechanisms are discussed below. 3.1.1 Erosion/Corrosion Erosion/corrosion is a flow accelerated form of corrosion due to the breakdown of a protective oxide layer on the surface of the piping. Several instances of carbon steel pipe wall thinning due to erosion/corrosion have been documented, but there is no history of wall thinning due to erosion/corrosion of stainless steel piping at nuclear power plants. Austenitic stainless steel is resistant to wall thinning by erosion/corrosion. The main steam and feedwater piping in the NPM is fabricated from SA-312 and SA-182 Type 304/304L (dual certified) austenitic stainless steel material and compatible austenitic stainless steel weld filler metals. The materials, in 2 3.6-37 Revision 4

The secondary water chemistry monitoring and control program described in Section 10.3.5 ensures that chloride, oxygen, fluoride, and sulfate levels do not cause erosion/corrosion in austenitic stainless steel in the main steam and feedwater piping. Stainless steel piping and components, such as letdown orifices, are potentially susceptible to erosion by cavitation under specific RCS flow conditions. Cavitation erosion has been observed in stainless steel piping in chemical and volume control systems of PWRs downstream of letdown orifices. Piping downstream of valves that significantly drop the pressure of the fluid in the system are also possible locations of cavitation erosion. The main steam and feedwater piping inside the CNV do not have inline components that significantly decrease the pressure of the fluid in the piping in the direction of flow. Therefore, conditions conducive to fluid cavitation do not exist. Based on the above discussion, erosion/corrosion induced wall thinning is not an issue for the main steam and feedwater piping subject to LBB. 3.1.2 Stress Corrosion Cracking If any one of the following three conditions is not present, stress corrosion-cracking (SCC) does not take place. The three conditions are:

There must be a corrosive environment.
The material itself must be susceptible.
Tensile stresses must be present in the material.

The main steam and feedwater piping is not susceptible to SCC because the piping is not exposed to a corrosive environment, the material is SCC resistant, and tensile stresses that could initiate SCC are not present. The secondary water chemistry monitoring and control program described in Section 10.3.5 ensures that chloride, oxygen, fluoride, and sulfate levels do not cause SCC in austenitic stainless steel in the main steam and feedwater piping. During reactor shutdown conditions, the outside surfaces of some piping inside the CNV are exposed to borated water. Minimizing the chloride levels in the water along with the low levels of oxygen in the water reduces the potential for SCC. The temperature of the water on the outside of the piping is maintained near room temperature, which prevents SCC initiation in conjunction with minimizing chlorides in solution. Water chemistry conditions during shutdown conditions are controlled to preclude SCC initiation from the outer surface of the piping, using water treatment methods discussed in Section 10.3.5. 2 3.6-38 Revision 4

weight percent carbon, which mitigates sensitization. The use of cold worked austenitic stainless steels is generally avoided; however, if used, the cold worked LBB pipes are followed by a solution annealing process. Based on the above, the LBB piping is not susceptible to SCC. 3.1.3 Creep and Creep Fatigue The design temperature for the MSS and FWS lines is 650 degrees F and normal operating temperatures are 585 degrees F and 300 degrees F respectively. Creep and creep fatigue are not a concern for austenitic steel piping below 800 degrees F. Because the design and operating temperatures of the piping systems are below these limits, creep and creep fatigue are not a concern. 3.1.4 Water Hammer/Steam Hammer The potential for water hammer and relief valve discharge loads are considered and their effects minimized in the design of the main steam system. Utilizing drain pots, proper line sloping, and drain valves minimize this potential. The dynamic loads such as those caused by main steam isolation valve closure or Turbine Stop Valve closure due to water hammer and steam hammer are analyzed and accounted for in the design and analysis of the main steam piping. Therefore, the main steam piping is not susceptible to effects of water hammer. The FWS and SG contain design features and operating procedures that minimize the potential for and effect of water hammer. The SG and FWS features are designed to minimize or eliminate the potential for water hammer in the steam generator FWS. The dynamic loads such as those caused by feedwater isolation valve closure and turbine trip due to water hammer are analyzed and accounted for in the design and analysis of the FWS piping. Therefore, the feedwater system LBB piping is not susceptible to water hammer. The safe shutdown earthquake loading used for the LBB evaluations bounds the water hammer loading for both the feedwater lines and the main steam lines. 3.1.5 Fatigue Low-cycle Fatigue The main steam and feedwater piping inside the CNV is ASME Class 2. Class 2 piping systems incorporate stress range reduction factors in accordance with Subsection NC of Section III of the ASME BPVC to account for cyclic loading. The reduction factors mitigate the need for a detailed fatigue evaluation including the calculation of cumulative usage factors. This design requirement ensures the piping is not susceptible to low-cycle fatigue due to operational transients. Confirmation is to be provided in the pre-operational thermal expansion monitoring program. 2 3.6-39 Revision 4

Main steam and feedwater piping design requirements also ensure the piping is not susceptible to high-cycle fatigue due to vibration. The main steam and feedwater lines are part of the NuScale Power Module and are included within the scope of the NuScale CVAP, see Section 3.9.2. Piping systems that meet the screening criteria for applicable flow induced vibration mechanisms are evaluated in the analysis program. If a large margin of safety is not demonstrated, prototype testing is performed in accordance with the CVAP measurement program. 3.1.6 Thermal Aging Embrittlement No cast steel is used for the main steam and feedwater piping. Wrought austenitic stainless steel is used. This product form is not susceptible to thermal aging embrittlement at the maximum design temperature of the piping. To minimize thermal aging embrittlement in austenitic stainless steel welds, delta ferrite content is controlled using the methods in RG 1.31. Delta ferrite for austenitic stainless steel weld filler metals with low molybdenum content, such as Type 308/308L, is limited to 5FN to 20FN. Delta ferrite for austenitic stainless weld filler metals with higher molybdenum content, such as Type 316/316L, is limited to 5FN to 16FN. 3.1.7 Thermal Stratification Thermal stratification in piping occurs when fluid at a significantly different temperature is introduced into a long horizontal run of piping. The main steam and feedwater lines inside the CNV do not have long horizontal runs and are therefore not susceptible to thermal stratification. 3.1.8 Irradiation Effects The main steam and feedwater piping materials, including austenitic stainless steels and compatible stainless steel welds, are not susceptible to irradiation embrittlement at the radiation levels outside the reactor vessel. The main steam and feedwater piping is not susceptible to Irradiation Assisted Stress Corrosion Cracking (IASCC) due to its low fluence. IASCC typically affects components such as core support structures in regions with high fluence, near the core and inside the reactor vessel. Because the main steam and feedwater piping is outside of the reactor vessel and above the core, the fluence is insufficient to be an IASCC concern. 3.1.9 Rupture from Indirect Causes The main steam and feedwater lines subject to LBB analysis are located inside the CNV. Rupture by indirect causes (e.g., fires, missiles, or natural phenomena) is precluded by design.

The NPM and the components inside the CNV are safety-related and Seismic Category I, this precludes adverse interactions from a seismic event.

2 3.6-40 Revision 4

There are no internal missile sources inside containment (see Section 3.5).
Containment is flooded as part of the normal shutdown process, therefore flooding is considered in the design.

3.1.10 Cleavage Type Rupture Cleavage type ruptures are not a concern for the main steam and feedwater lines. Austenitic stainless steel is highly ductile and resistant to cleavage type ruptures at system operating temperatures and the lower temperatures experienced during shutdown conditions. 3.2 Materials The MSS and FWS piping is fabricated from SA-312 and SA-182 TP304/TP304L (dual certified) material. Alloy 600 and weld metal Alloy 82/182 are not used in the NPM LBB piping discussed. 3.2.1 Geometry The main steam piping is evaluated in six segments: Section Geometry Nominal Nominal Inside Thickness t, (in.) Diameter (in.) NPS 8, SCH 120 straight and curved pipe base metal 7.187 0.719 NPS 8, SCH 120 pipe-to-pipe weld 7.187 0.719 NPS 8, SCH 120 pipe-to-safe-end weld 7.187 0.719 NPS 12, SCH 120 straight and curved pipe base metal 10.75 1.000 NPS 12, SCH 120 pipe-to-safe-end weld 10.75 1.000 NPS 8, SCH 120 elbow base metal 7.187 0.719 The feedwater piping is evaluated in four segments: Section Geometry Nominal Nominal Inside Thickness t, (in.) Diameter (in.) NPS 5, SCH 120 straight and curved pipe base metal 4.563 0.500 NPS 5, SCH 120 pipe-to-pipe, pipe-to-tee, pipe-to-safe-end, 4.563 0.500 tee-to-tee welds NPS 4, SCH 120 straight and curved pipe base metal 3.624 0.438 NPS 4, SCH 120 pipe-to-tee pipe-to-safe-end welds 3.624 0.438 3.2.2 Operating Conditions and Load The operating pressure and temperature for the MSS piping are 500 psia and 585 degrees F, respectively. 2 3.6-41 Revision 4

3.2.3 Materials The MSS piping base metal is made of SA-312 and SA-182 Grade TP304/TP304L (dual certified). The pipe-to-pipe weld and pipe-to-safe-end weld are both made with austenitic stainless steel weld filler material. The tensile material properties used in the analysis of MSS materials are either at 550 degrees F or 585 degrees F. It is acceptable to use material properties at 550 degrees F to approximate the material properties at the actual operating temperature (585 degrees F) because the variations in the material properties between these temperatures are insignificant. The FWS piping base metal is made of SA-312 Grade TP304/TP304L. The pipe-to-pipe, pipe-to-safe-end, pipe-to-tee, tee-to-tee welds are made with austenitic stainless steel weld filler material. The tensile material properties used in the analysis of FWS materials are at 300 degrees F. Only gas tungsten arc welding is used for main steam and feedwater piping subject to LBB qualification and the weld filler metals are limited to the following:

SFA-5.9: ER308, ER308L, ER316, ER316L
SFA-5.30: IN308, IN308L, IN316, IN316L 3.2.4 Tensile Material Properties Material y (ksi) u (ksi) E (ksi) o n Main Steam Piping SA-312 TP304 18.7(1) 63.4(1) 25450(1) 0.00073(5) 8.07(4) 3.80(4)

ER308L Weld 22.1(7) 75.0(2) 25450 (1) 0.00087 (5) 2.31(3) 3.28(3) Feedwater Piping SA-312 TP304 22.4(1) 66.2(1) 27000(1) 0.00083(5) 2.411(3) 3.616(3) ER308L Weld 25.4 (6) 75.0(2) 27000(1) 0.00094(5) 2.126(3) 3.616(3) Notes (1) ASME Boiler and Pressure Vessel Code, Section II, Part D, 2013 Edition no Addenda. (2) ASME Boiler and Pressure Vessel Code, Section II, Part C, 2013 Edition no Addenda. (3) , n are R-O Model coefficient and exponent evaluated by method for elastic plastic fracture analysis that determines the R-O parameters (, n) from basic mechanical properties determined from the ASME Code. (4) from Reference 3.6-10 2 3.6-42 Revision 4

(6) The weld metal minimum yield strength is assumed to be 25.4 ksi at 300 degrees F. This value is obtained from the base metal yield strength ratioed up by the ratio of the weld metal minimum ultimate strength to the base metal minimum ultimate strength. (7) The weld metal minimum yield strength is assumed to be 22.1 ksi at 575 degrees F. This value is obtained from the base metal yield strength ratioed up by the weld metal minimum ultimate strength to the base metal minimum strength. 3.2.5 Crack Morphology Parameters For fatigue cracks in pipes, the crack morphology parameters are obtained from Tables 3.3 through 3.8 of NUREG/CR-6004, "Probabilistic Pipe Fracture Evaluations for Leak-Rate-Detection Applications," (Reference 3.6-10). The mean values are listed below: Parameter (Units) Mean Value Global roughness (inch) 1325 Local roughness (inch) 317 Number of 90-degree turns (inch-1 ) 64 Global path deviation 1.07 Global and local path deviation 1.33 3.3 Analysis Methodology To ensure that an adequate margin exists for leak detection, the analysis assumes a leak rate 10 times larger than the minimum plant leak detection capability. A margin of 2.0 on flaw size and a margin of 1.0 on load is used when using the algebraic sum load combination method as described in Section 3.6.3.3.1.1. Therefore, for a given flaw size that develops a detectable leakage with safety factor of 10, a fracture mechanics analysis is performed using twice the leakage flaw size to obtain a maximum allowable stress. The maximum allowable stress must be equal to or greater than the actual applied stress. 3.3.1 Load Combination Method It is allowable to use either the absolute sum load combination method or the algebraic sum load combination method, which require different margins on the flaw size. Both load combination methods consider deadweight (DW), thermal expansion (TH), flow loads due to pressure (PR), safe shutdown earthquake (SSE) inertial and seismic anchor motion (SAM) loads. 2 3.6-43 Revision 4

The axial force, F, and moment, M, can be algebraically summed if a margin factor SM of 1.4 is applied for the applicable DW, TH, PR, SSE, and SAM loads. F Combined = S M ( F DW + F TH + F PR + F SSE + F SAM ) Eq. 3.6-2 M i, Combined = S M ( M i, DW + M i, TH + M i, PR + M i, SSE + M i, SAM ) Eq. 3.6-3 Where FDW, FTH, FPR, FSSE and FSAM are axial force (with a unit of lbf) due to deadweight, thermal expansion, internal pressure, SSE and SAM, respectively, and Mi,DW, Mi,TH, Mi,PR, Mi,SSE, and Mi,SAM are moment (with a unit of in-lbf) due to deadweight, thermal expansion, internal pressure, SSE and SAM, respectively, for component i (i = X, Y, Z). SM is the safety margin for load combination. First, for the algebraic sum method of load combination, the margin SM is set to 1.4. If the allowable flaw length from the flaw stability analysis is at least equal to the leakage size flaw, then the margin on load is met. Second, the margin SM is set to 1.0 and if the allowable flaw length from the flaw stability analysis is at least twice the leakage size flaw, then the margin on flaw size is met. 3.3.1.2 Absolute Sum Method The loads can also be combined based on individual absolute values as follows: F Combined = F DW + F TH + F PR + F SSE + F SAM Eq. 3.6-4 M i, Combined = M i, DW + M i, TH + M i, PR + M i, SSE + M i, SAM Eq. 3.6-5 The total moment for the primary bending stress is calculated as square root of the sum of squares (SRSS): M Combined = M x2 , Combined + M y2 , Combined + M z2 , Combined Eq. 3.6-6 For an absolute sum load combination method, the margin on the load SM is set to 1.0. If the allowable flaw length from the flaw stability analysis is equal to at least twice the leakage size flaw, the margins on load and flaw size are met. 3.3.2 Piping Load Combination For normal stress calculation, the algebraic sum is used for load combinations based on SRP 3.6.3 paragraph III.11(c)(iii). The normal operating axial force and moments are calculated by the following equations: 2 3.6-44 Revision 4

M X = ( M X ) DW + ( M X ) TH Eq. 3.6-7 M Y = ( M Y ) DW + ( M Y ) TH M Z = ( M Z ) DW + ( M Z ) TH Where FDW, FTH, FPR, Mi,DW and Mi,TH (i = X, Y, Z) are defined in Section 3.6.3.3.1.1. The resultant moment is then calculated as the SRSS: M = MX2 + MY2 + MZ2 Eq. 3.6-8 For the maximum stress calculation, the maximum axial force and moments are: F = F DW + F PR + F SSE M X = ( M X ) DW + ( M X ) SSE Eq. 3.6-9 M Y = ( M Y ) DW + ( M Y ) SSE M Z = ( M Z ) DW + ( M Z ) SSE Where Mi,SSE (i = X, Y, Z) are defined in Section 3.6.3.3.1.1. The resultant moment is then calculated as the SRSS: M = MX2 + MY2 + MZ2 Eq. 3.6-10 In the above equations, the moment due to the internal pressure is not included although it is included in Eq. 3.6-3 and Eq. 3.6-5, because the moment due to internal pressure is negligible. For limit load analysis, the thermal expansion and SAM loads are not included in Eq. 3.6-51 because they are secondary loads. The stresses due to axial loads and moments are then calculated by:

                                      = F         M
                                             --- + -----                       Eq. 3.6-11 A Z where, A = cross-sectional area, Z = section modulus, M = moment, and F = axial force.

2 3.6-45 Revision 4

3.3.3.1 Elastic-Plastic Fracture Mechanics Methods The first step of the leakage rate calculation is to determine the crack opening area, based on elastic-plastic fracture mechanics methods. Although finite element method and computational fracture mechanics can be used to calculate crack opening displacement and crack opening area, it is computationally inefficient when applied for LBB, because many iterations may be needed to find the crack size and the crack opening displacement to produce a detectable leakage rate, or bounding analysis curves may need to be developed. The GE/EPRI method (Reference 3.6-14) is used in this LBB calculation because it is easier to implement and is validated by experimental data. The GE/EPRI method was developed for three loading conditions: pure tension, pure bending, and combined tension and bending. The crack opening displacement includes an elastic portion and a perfectly-plastic portion based on a Ramberg-Osgood (R-O) material model in Eq. 3.6-12. n

                                         ----- = ------ +   ------

0 0 0 Eq. 3.6-12 where,

          = true strain, 0 = reference strain (given by ---- ),

E E = Youngs modulus (psi),

          = true stress (psi),

0 = reference stress (the ASME Code-specified 0.2 percent offset yield strength y in this calculation) (psi), and

         , n = R-O model coefficient and exponent.

3.3.3.1.1 Crack Opening Displacement for Through-Wall Cracks in Cylinders under Remote Bending In the linear elastic range, the elastic crack opening displacement e of the total mouth opening displacement of a pipe, as illustrated in Figure 3.6-19, due to a remote bending stress can be expressed as: 4 B a B a R e = ------------- V --- , --- Eq. 3.6-13 E 1b t 2 3.6-46 Revision 4

a = Rm = half crack length at the mean radius, Eq. 3.6-14 b = R m = half pipe circumference, Eq. 3.6-15

  = half crack angle in radians, R = mean pipe radius, R = R m ,

E = modulus of elasticity. B = MR--------- = remote bending stress Eq. 3.6-16 I M = remote bending moment. 1 I = --- R - R R t 4 4 3

                                                = area moment of inertia Eq. 3.6-17 4         0       i R0, Ri = pipe outer and inner radius, t = pipe wall thickness, and B

V = influence function for elastic crack opening displacement under 1 bending, given as tabulated values for various crack sizes and pipe geometries in Table 6-5 of Reference 3.6-2 for straight pipe, and in Tables F.1 and F.2 of Reference 3.6-5 for elbows. It is noted that a--- = --- , so they are used interchangeably. b The plastic portion of crack opening displacement is expressed as: B a R M n p = 0 aH --- ,n, --- ------- Eq. 3.6-18 2b t M 0 where,

 , n = R-O model coefficient and exponent, and B

H = influence function for plastic crack opening displacement under 2 bending, given as tabulated values for various crack sizes, material R-O model exponents, and pipe geometries in Tables 6-6, 6-7, and 6-8 of Reference 3.6-2 for straight pipe, and in Tables F.1 and F.2 of Reference 3.6-5 for elbows. 2 3.6-47 Revision 4

A discussion of -correction is presented in Section 3.6.3.3.3.1.3 The total crack opening displacement is then calculated by 4 B a B a R B a R M m

                   =  e +  p = ------------- V  --- , --- +  o aH  ---, n, ---  -------      Eq. 3.6-20 E         1 b t                            2b      t   M 0 3.3.3.1.2 Approach to Handle Combined Axial Force and Bending Moment To apply the influence functions from the bending condition to combined tension and bending, the axial force can be converted to an equivalent bending moment and added to the applied moment. The stress intensity factors due to axial force and bending moment can be expressed as:

F K T = ------------ aF T ( ) Eq. 3.6-21 2Rt M K B = -----------2 aF B ( ) Eq. 3.6-22 R t where, 3 5 7 2 2 2 F T ( ) = 1 + 7.5 --- - 15 --- + 33 --- Eq. 3.6-23 3 5 7 2 2 2 F B ( ) = 1 + 6.8 --- - 13.6 --- + 20 --- Eq. 3.6-24 Note that the equations are derived for R/t=10. It is expected that the approximation is acceptable for R/t between 5 and 20. The equivalent moment due to an axial force P is then calculated by: FR F T ( ) M e = ------- --------------- Eq. 3.6-25 2 FB ( ) 3.3.3.1.3 - Correction to the Crack Opening Displacement Models In Reference 3.6-6, the improved crack opening displacement estimation scheme is proposed to better match the GE/EPRI estimation to the experimental data. For pure bending or tension, the plastic part of the crack opening displacement is given below. 2 3.6-48 Revision 4

1 B a R M n p = 0 aH --- ,n, --- ------- n Eq. 3.6-26 2 b t M 0 For pure tension 1 T a R F n p = 0 aH --- ,n, --- ------ n Eq. 3.6-27 2 b t F 0 Here, is replaced by the term 1/n. Because is normally greater than 1, the effect of this term is to reduce the crack opening displacement relative to what would be computed using Eq. 3.6-18. A different correction is needed for the combined tension and bending case because the plastic contributions from pure tension and pure bending cannot be added linearly. For a simplified approximation, the following is used: 1n B a R M n p = 0.5 ( + ) 0 aH 2 --- ,n ,--- -------- Eq. 3.6-28 b t M 0 The -correction in Eq. 3.6-27 is applied when using the bending influence function with the equivalent moment calculated by Eq. 3.6-25. 3.3.3.1.4 Crack Opening Area and Hydraulic Diameter The crack opening profile is assumed to be elliptical. The crack opening area is calculated by: A crack = a/2 Eq. 3.6-29 The perimeter of an ellipse can be approximated by P wetted [ 3 ( a + /2 ) - ( 3a + /2 ) ( a + 3/2 ) ] Eq. 3.6-30 The hydraulic diameter is then calculated by 4A D H = ----------------- Eq. 3.6-31 P wetted The crack opening area and the hydraulic diameter are two major crack geometric parameters that are needed for leak rate analysis, as presented in Section 3.6.3.3.3.2. 2 3.6-49 Revision 4

The Henry-Fauske thermal-hydraulic model of two-phase flow (Reference 3.6-8, Reference 3.6-9, and Reference 3.6-10) through long channels, as illustrated in Figure 3.6-20, forms the basis for the leak rate analysis. Compared to other simplified homogenous models, this model is a slip-flow model in the sense that the vapor has a higher velocity than the liquid in the vapor-liquid mixture of a two-phase flow system. A slip ratio, defined as the ratio of gas velocity to liquid velocity, is used in the homogeneous equilibrium model equations. When the two-phase mixture experiences critical flow, the time required for the fluid to reach thermodynamic equilibrium when moving into regions of lower pressure is comparable to the time that the fluid is flowing in the crack, which leads to non-equilibrium vapor generation rates for two-phase critical flows. To account for these non-equilibrium effects, Henry and Fauske assumed that the mixture quality relaxes in an exponential manner toward the equilibrium quality that would be obtained in a long tube. The relaxation coefficient was calculated based on their experiments with the critical flow of a two-phase water mixture in long tubes, with the ratio of flow-path length to pipe inside diameter greater than 100. 3.3.3.2.1 Thermal-hydraulic Model of Two-phase Flow In the LBB analysis, the Henry-Fauske model of two-phase flow through long channels is applied to calculate leak rates. Mass flux equilibrium is written in the following format:

                                                                                             -1 2          x c v gc                      dx e
                               = G c - ------------           - (  gc -  lc )N 1 --------    = 0 Eq. 3.6-32 o Pc                          dP Subject to the constraint in terms of pressure equilibrium
                            = P c + P e + P f + P a + P aa + P k - P o = 0                    Eq. 3.6-33 where, Gc   = mass flux of the fluid at the crack exit plane, c

So - Sl x e = -------------------

                                                     -   = equilibrium fluid quality                Eq. 3.6-34 c           c Sg - Sl So  = entropy at entrance of the crack plane, c

Sl = entropy of the saturated liquid at the crack exit plane pressure, 2 3.6-50 Revision 4

20x e , if x e <0.05 N1 = Eq. 3.6-35 1.0, if x e 0.05 L

                                                                               - B  ------a- - 12 DH xc = N1 xe 1 - e                                     Eq. 3.6-36 La = flow-path length, 4  Crack Opening Area D H = ------------------------------------------------------------
                                                                  - = the hydraulic diameter perimeter (see Crack Opening Perimeter Eq. 3.6-31),

B=0.0523 = a constant based on experiments used in calculating exponential mixture quality relaxation, vgc = specific volume of saturated vapor at exit pressure, vlc = specific volume of saturated liquid at exit pressure, o = isentropic expansion exponent, P = pressure, Pc = absolute pressure of the fluid at the crack exit plane, P0 = absolute pressure at the entrance of the crack plane, 2 G v lo o P e = -------------- = pressure loss due to entrance effects Eq. 3.6-37 2 2C D Go = mass flux of the fluid at the crack entrance plane, vlo = specific volume of the saturated liquid at the entrance pressure, CD = discharge coefficient. A value of 0.95 is recommended for tight cracks, 2 La G c P f = f ----- ------- [ ( 1 - x )v l + xv g ] = Pressure loss due to friction Eq. 3.6-38 Di 2 x = average fluid quality, 2 3.6-51 Revision 4

v l = average specific volume of saturated liquid, DH -2 f = 2log ------- + 1.74 = Von Karman friction factor Eq. 3.6-39 2

          = crack face roughness, 2

P a = G [ ( 1 - x c )v lc + x c v gc - v lc ] = pressure loss due to T acceleration of the fluid as it flows through the crack Eq. 3.6-40 G T = average mass flux in the two-phase region of crack flow, P aa = acceleration pressure loss due to area change is assumed zero, 2 G P k = e v ------ [ v l + x ( v g - v l ) ] = pressure loss due to ends and 2 protrusions Eq. 3.6-41 2 G = average mass flux G2 of the fluid e v = e n L a = the total loss coefficient over the flow path Eq. 3.6-42 en = the number of velocity heads lost per unit flow path length, which is given in Eq. 3.6-44. Eq. 3.6-33 and Eq. 3.6-32 are evaluated by iteration to give the leak flow rate through the crack and the exit pressure for given crack inlet stagnation conditions and crack geometry. 3.3.3.2.2 Effective Crack Morphology Parameters In NUREG/CR-6004 (Reference 3.6-10), a modified model was developed to define the surface roughness, effective flow path length and the number of turns as a function of the ratio of the crack opening displacement () to the global roughness (G) of the flow path, which is considered to be more realistic. The basic idea is depicted in Figure 3.6-21. For a very tight crack, i.e., / G < 0.1 , the effective roughness is close to the local roughness (L). But for a crack with wide opening, i.e., / G > 10 , the effective roughness is close to the global roughness. A linear function is used to calculate the effective roughness in between. The effective roughness, , is then expressed as 2 3.6-52 Revision 4

0 < ------- < 0.1 L G G - L

                                              - ------- - 0.1 ,
                  =   L + ------------------                             0.1  -------  10          Eq. 3.6-43 9.9   G                                           G
                                                                            ------- > 10 G                                                   G Similarly, for a very tight crack, i.e., / G < 0.1 , the effective number of turns is close to the number of local turns. But for a crack with wide opening, i.e., / G > 10 , the effective number of turns decreases to about 10 percent of the local number of turns ( e n ). A linear function is used to L

calculate the effective number of turns in between. The effective number of turns is then expressed as en , 0 < ------- < 0.1 L G en L e n = e - ------ 0.1 ------- 10 Eq. 3.6-44 n L 11- ------- - 0.1 , G G 0.1e , nL ------- > 10 G In a similar way, the actual crack path to thickness ratio that represents the correction factor for flow path deviation from straightness is also a function of crack opening displacement. For a very tight crack, i.e., / G < 0.1 , the effective deviation is close to the global plus local path deviation K G + L . But for a crack with wide opening, i.e., / G > 10 , the effective deviation is close to the global path deviation K G . A linear function is used to calculate the effective deviations in between. The effective deviation factor is then expressed as: K

                           ,                                                         0 < ------- < 0.1 G+L                                                                      G La                   KG + L - KG                                                   
                                                        - ------- - 0.1 ,
          ----- =  K G + L - ---------------------------                            0.1  -------  10 Eq. 3.6-45 t                           9.9                                                G G

K ,

                                                                                     ------- > 10 G                                                                G These crack opening displacement-dependent effective crack morphology parameters are plotted in Figure 3.6-22.

2 3.6-53 Revision 4

The leakage of the piping systems inside the CNV can be detected by either using the CNV pressure sensor or the CES sample vessel instrumentation. See Section 3.6.3.5 for more discussion. The minimum detectable leak rate is 0.01 lbm/min, or 0.001 gallon per minute (GPM). Per SRP 3.6.3, a safety margin of 10 is required for the detectable leak rate. However, a more conservative leak rate of 0.2 lbm/min (or 2.0 lbm/min after the margin of 10 is applied) is used as the leak rate to construct the LBB bounding curves. 3.3.4 Flaw Stability Analysis Method (Limit Load Analysis) It is required that any subcritical cracks, including surface and through-wall cracks in circumferential and axial directions be stable so that a catastrophic break is not possible. The cracks in an elbow also need to be evaluated if not bounded by the straight piping. Crack growth evaluation is required to be performed to ensure that cracks are stable. It is usually found that circumferential through-wall cracks are more limiting than axial or surface cracks. Because the LBB analysis is performed for austenitic stainless steel piping systems, the stability assessment is based on limit load analysis. A modified limit load analysis based on the master curve is used to calculate the allowable stable flaw size. The master curve is constructed to be a stress index S I as a function of the postulated total circumferential through-wall flaw size 2a c . The stress index S I and the half flaw size a c are expressed as: 2

                              --------f ( 2 sin  - sin  ) + ( S m ) ( P m ), if  +

SI = Eq. 3.6-46 2 f

                              -------- sin  + ( S m ) ( P m ), if  +  >

where, P m 0.5 ( - ) - ---------- , if + f

                               =                                                           Eq. 3.6-47 P m
                                            - ---------- , if  +  >

f F P m = -----x = primary membrane stress Eq. 3.6-48 A Fx = total applied axial force, A = cross-section area, 2 3.6-54 Revision 4

Rm Rm = pipe mean radius, SM = 1 = safety margin on the load, f = 0.5 ( y + u ) = flow stress Eq. 3.6-50 y = yield strength, and u = ultimate strength. The stress index is also expressed in SRP 3.6.3 as: SI = SM ( Pm + Pb ) Eq. 3.6-51 where, M Rm P b = ---------------- = primary bending stress Eq. 3.6-52 I F

                                          - x I max -----         A M = -------------------------------- = applied maximum moment Eq. 3.6-53 Rm max  = applied maximum stress, and I = area moment of inertia.

The max can be determined by making SI in Eq. 3.6-46 equal to that in Eq. 3.6-51. 3.3.5 Development of Smooth Bounding Analysis Curve To develop a smooth bounding analysis curve (SBAC), the following steps are used:

1) prepare the required inputs as discussed in Geometry and Material Properties Section 3.6.3.2.1 and Section 3.6.3.2.4, and Normal Loads Section 3.6.3.3.2

2) low normal stress case - calculate the axial force for normal operating pressure and the bending moment based on a selected lower magnitude of bending stress that is lower than the expected minimum bending stress

3) calculate the leakage flaw size at 100 percent power condition for 10 times the leak detection capability using the methodology discussed in Section 3.6.3.3.3

4) perform the stability analysis using the limit load methodology for austenitic stainless steel piping discussed in Section 3.6.3.3.4. The maximum bending 2 3.6-55 Revision 4

5) calculate the low normal stress and corresponding maximum stress using the axial force and the bending moments by Eq. 3.6-11 to establish the first point on the SBAC

6) high normal stress case - calculate the axial force for normal operating pressure and the bending moment based on a selected higher magnitude of bending stress that is close to the material flow stress. Calculate the corresponding maximum stress following Steps 3 through 4

7) establish the last point on the SBAC for the High Normal Stress Case following Steps 3 through 6

8) determine intermediate points along the abscissa by equal division of abscissa points between the first and the last points

9) calculate the intermediate points following Steps 3 through 5

10) develop the SBAC by joining these points to form a smooth curve 3.3.6 Application of SBACs The SBACs are used during the design of the piping systems to provide a design that satisfies LBB criteria. In addition, the results of the piping analysis are reconciled to the SBACs to verify that the fabricated piping systems satisfy LBB criteria. To evaluate the LBB applicability, the results of the pipe stress analysis are compared to the applicable SBAC at the critical location with highest maximum stress. At critical locations, the load combination for the normal stress and maximum stress calculation uses the methods presented in Section 3.6.3.3.2. The procedure for LBB analysis discussed in this section is illustrated by a flow chart shown in Figure 3.6-18.

3.4 Analysis of Main Steam and Feedwater Piping inside Containment 3.4.1 Analysis of Main Steam Piping Based on piping materials (base and weld metal) and configurations (pipe and elbow) in Section 3.6.3.2.1, six sections are analyzed. For each analysis, the piping stresses are determined based on the equations in Section 3.6.3.3.2. The SBAC are developed by first performing the limit load analysis to estimate the critical crack size based on Section 3.6.3.3.4. The half critical crack size is then used in the leakage rate analysis that builds in a safety margin of 2 on the crack size. The crack opening area is assumed to be constant through the thickness. The crack opening displacement is calculated using elastic-plastic fracture mechanics following Section 3.6.3.3.3. Plastic zone correction is not applied. Finally, the piping stresses and SBAC are compared to see if the pipe qualifies for LBB. 2 3.6-56 Revision 4

3.4.1.1.1 Normal Stress and Maximum Stress This analysis is for straight and curved NPS 8 pipes. Various locations in both main steam lines 1 and 2 are considered in this analysis. For each location, the normal stress and maximum stress are calculated using the equations in Section 3.6.3.3.2. By using Eq. 3.6-7 and Eq. 3.6-8, the normal axial force and moment are calculated. The maximum axial force and moment are calculated using Eq. 3.6-9 and Eq. 3.6-10. Lastly, the axial end cap force due to the internal pressure is added to the normal and maximum axial forces for calculating stress using Eq. 3.6-11. The resultant normal and maximum stresses for the main steam lines 1 and 2 locations are plotted (legends MS1 and MS2) in Figure 3.6-23. 3.4.1.1.2 SBAC Development The limit load analysis is performed first to estimate the critical crack size based on methodology described in Section 3.6.3.3.4. Half of the critical crack size is then used in leakage rate analysis. The crack opening displacement calculation using elastic-plastic fracture mechanics is based on the methodology discussed in Section 3.6.3.3.3. The leakage rate is calculated for the half critical crack size, which results in a leakage rate of 2.0 lbm/min, based on the detectable leak rate discussed in Section 3.6.3.3.3.3. Following the steps in Section 3.6.3.3.5, more points with higher normal stress are established for developing SBAC. The resultant SBAC is illustrated in Figure 3.6-23. It is observed that the stress points are below the SBAC, demonstrating the analyzed section satisfies LBB criteria. 3.4.1.2 NPS 8 Pipe-to-Pipe Weld This analysis is for circumferential welding between NPS 8 pipe and NPS 8 pipe. All NPS 8 pipe-to-pipe weld locations in both MS lines 1 and 2 are considered in this analysis. Following the same method described in Section 3.6.3.4.1.1, the normal and maximum stresses are calculated for each location in NPS 8 pipe-to-pipe weld. The resultant stresses are plotted in Figure 3.6-24. The SBAC is developed using the same method described in Section 3.6.3.4.1.1.2, except the weld material properties used are for ER308L. Using the methodology discussed in Section 3.6.3.3.3 for the COD calculation, the resultant SBAC is illustrated in Figure 3.6-24. It is observed that the stress points are below the SBAC, demonstrating the analyzed section satisfies LBB criteria. 2 3.6-57 Revision 4

This analysis is for circumferential welding between NPS 8 pipe and a safe end. All NPS 8 pipe-to-safe-end locations in both main steam lines 1 and 2 are considered in this analysis. The calculated normal and maximum stresses are plotted in Figure 3.6-25. The SBAC for NPS 8 pipe-to-safe-end weld is identical to that for NPS 8 pipe-to-pipe weld because their weld material and dimensions are identical. The SBAC chart, illustrated in Figure 3.6-25, shows that the stress points are below the SBAC, demonstrating the analyzed section satisfies LBB criteria. 3.4.1.4 NPS 12 Straight Pipe Base Metal This analysis is for straight and curved NPS 12 pipes. Various locations in both main steam lines 1 and 2 are considered in this analysis. The calculated normal and maximum stresses are plotted in Figure 3.6-26. For developing SBAC, the methodology discussed in Section 3.6.3.3.3 is used to calculate crack opening displacement. The resultant SBAC is illustrated in Figure 3.6-26. It is observed that the stress points are below the SBAC, demonstrating the analyzed section satisfies LBB criteria. 3.4.1.5 NPS 12 Pipe-to-Safe-End Weld This analysis is for circumferential welding between a NPS 12 pipe and a safe end. All NPS 12 pipe-to-safe-end weld locations in both MS lines 1 and 2 are considered in this analysis. The calculated normal and maximum stresses are plotted in Figure 3.6-27. For developing SBAC, the methodology discussed in Section 3.6.3.3.3 is used to calculate crack opening displacement. The resultant SBAC is illustrated in Figure 3.6-27. It is observed that the stress points are below the SBAC, demonstrating the analyzed section satisfies LBB criteria. 3.4.1.6 NPS 8 Elbow Base Metal This analysis is for NPS 8 elbows. Various locations in both MSS lines 1 and 2 are considered in this analysis. The calculated normal and maximum stresses are plotted in Figure 3.6-28. The resultant SBAC is illustrated in Figure 3.6-28. Note that the SBAC is developed by only four points because the V1 parameters become negative with higher normal stresses. This is due to the fact that the available parameters are for =45° and 90°, while the calculated beyond the fourth point is away from that range. Therefore, the calculated results beyond the fourth point are not considered. However, the trend of the four points in SBAC shows that the stress points are below the SBAC, demonstrating the analyzed section satisfies LBB criteria. 2 3.6-58 Revision 4

Based on piping materials (base and weld metals) and geometric parameters in Section 3.6.3.2.1, four sections are analyzed. For each analysis, the piping stresses are determined based on the equations in Section 3.6.3.3.2. The SBAC are developed by first performing the leak rate analysis based on Section 3.6.3.3.3 to estimate the leakage crack size that produces a leak rate equal to 10 times the minimum detectable leak rate. The leakage crack size is then used as the half critical crack size in the limit load analysis, based on Section 3.6.3.3.4, building in a safety margin of 2 on the crack size. The crack opening displacement is calculated using elastic-plastic fracture mechanics following Section 3.6.3.3.3. Plastic zone correction is used for the purpose of H2 B function calculation for the NPS 4 FWS lines, to be consistent with the method in Reference 3.6-2. Finally, the piping stresses and SBAC are compared to confirm that the pipe qualifies for LBB. 3.4.2.1 Normal and Maximum Stress Calculations For each location considered, the normal stress and maximum stress are calculated using the equations in Section 3.6.3.3.2. By using Eq. 3.6-7 and Eq. 3.6-8, the normal axial force and moment are calculated. The maximum axial force and moment are calculated using Eq. 3.6-9 and Eq. 3.6-10. Lastly, the axial end cap force due to the internal pressures is added to the normal and maximum axial forces for calculating stress using Eq. 3.6-11. 3.4.2.2 NPS 4 Feedwater System Line Base Metal Various locations in both FWS lines 1 and 2 are considered in the analysis for straight and curved NPS 4 pipe base metal. For each location, the normal stress and maximum stress are calculated using the equations in Section 3.6.3.3.2, following the method described in Section 3.6.3.4.2.1. The resultant normal and maximum stresses for the locations are then plotted (legends FWS Line 1 and FWS Line 2) in Figure 3.6-29, the SBAC Chart for NPS 4 FWS line base metal. The SBAC is developed using the method described in Section 3.6.3.3.5. The stress points are below the SBAC, demonstrating that the analyzed section satisfies the LBB criteria. 3.4.2.3 NPS 4 Feedwater System Line Welds The analysis addressed the circumferential welds including pipe-to-tee, and pipe to safe-end welds. All NPS 4 weld locations in both FWS lines 1 and 2 were considered. Following the same method described in Section 3.6.3.4.2.1, the normal and maximum stresses were calculated for each location of the NPS 4 line welds. The resultant stresses are plotted in Figure 3.6-30, the SBAC Chart for NPS 4 FWS line welds. 2 3.6-59 Revision 4

that the analyzed section satisfies the LBB criteria. 3.4.2.4 NPS 5 Feedwater System Line Base Metal Various locations in both FWS lines 1 and 2 are considered in the analysis for straight and curved NPS 5 pipe base metal. Following the same method described in Section 3.6.3.4.2.1, the normal and maximum stresses are calculated for each location in the NPS 5 base metal. The calculated normal and maximum stresses are plotted in Figure 3.6-31, the SBAC Chart for NPS 5 FWS line base metal. The SBAC is developed using the method described in Section 3.6.3.3.5 and plotted in Figure 3.6-31. The stress points are below the SBAC, demonstrating that the analyzed section satisfies the LBB criteria. 3.4.2.5 NPS 5 Feedwater System Line Welds The analysis addressed the circumferential welds including pipe to tee, and pipe to safe end welds. All NPS 5 weld locations in both FWS lines 1 and 2 were considered. Following the same method described in Section 3.6.3.4.2.1, the normal and maximum stresses were calculated for each location of the NPS 5 line welds. The resultant stresses are plotted in Figure 3.6-32, the SBAC Chart for NPS 5 FWS line welds. The SBAC is developed using the method described in Section 3.6.3.3.5 and plotted in Figure 3.6-32. The stress points are below the SBAC, demonstrating that the analyzed section satisfies the LBB criteria. 3.4.3 Results and Conclusions 3.4.3.1 Main Steam System Piping The LBB allowable maximum axial and bending stress loads are compared against the actual normal operating plus SSE loadings of the MSS piping. The actual loads (the combined axial loads and the combined bending stresses as defined in SRP 3.6.3), for a given LBB location, fall within the SBAC depicted in Figure 3.6-23, Figure 3.6-24, Figure 3.6-25, Figure 3.6-26, Figure 3.6-27 and Figure 3.6-28. Therefore, it is concluded that the MSS piping meets the LBB criteria. 3.4.3.2 Feedwater System Piping The LBB allowable maximum axial and bending stress loads are compared against the actual normal operating plus SSE loadings of the FWS piping. The actual loads (the combined axial loads and the combined bending stresses as defined in SRP 3.6.3), for a given LBB location, fall within the SBAC depicted in 2 3.6-60 Revision 4

3.5 Leak Detection Section 5.2.5 describes the leak detection system for inside the CNV. The SRP 3.6.3 states "The specifications for plant-specific leakage detection systems inside containment are equivalent to those in Regulatory Guide 1.45." As noted in Section 5.2.5, the reactor coolant pressure boundary leakage detection systems for the NPM conform to the sensitivity and response times recommended in RG 1.45, Revision 1. This section describes the analysis methods used to support the application of LBB to high-energy piping in the NPM. Regulatory Guide 1.45 Regulatory Position 2.1 states plant procedures should include the collection of leakage to the primary reactor containment from unidentified sources so that the total flow rate can be detected, monitored, and quantified for flow rates greater than 0.05 gpm. According to RG 1.45 Regulatory Position 2.2, the plant should use leakage detection systems with a response time of no greater than 1 hour for a leakage rate of 1 gpm. Leakage monitoring is provided by two means, change in pressure within the CNV and collected condensate from the CES sample vessel. The minimum detectable leak rate for the CES sample vessel is not easily quantified, because all liquid or vapor leaks within the CNV are eventually collected in the CES sample vessel. Once in the CES sample vessel, the minimum detectable volume is 0.042 gal or 0.333 lb of liquid. While there is theoretically no minimum detectable leak rate, main steam and feedwater system leak rates of 0.001 gpm or 0.01 lbm/min take less than 60 minutes to accumulate more than the minimum detectable volume. To satisfy Regulatory Position 2.1 of RG 1.45, once the operators observe a pressure change in containment, a leak rate procedure is initiated to quantify the total leak rate. This, combined with other indications can aid in determining the leak source. In this instance, leaks can be detected using the CES sample vessel, where condensable fluids are collected after they are removed from containment via the vacuum pumps. The sample vessel level is configured to alarm the control room. Once a higher equilibrium pressure is reached during a leak scenario, leak rate measurements can be taken with the CES alone, using the CES sample tank. 4 High Energy Line Break Evaluation (Non-LBB) 4.1 Postulation of Pipe Breaks in Areas Other than Containment Penetration Where break locations are selected without the benefit of stress calculations, breaks are postulated at the piping welds to each fitting, valve, or welded attachment. Breaks in non-ASME Class piping are addressed in Section 3.6.2.1.8. Additionally, in accordance with BTP 3-4, Part B, Item A(iii)(4), if a structure is credited with separating a high-energy line from an essential SSC, that separating structure is designed to withstand the consequences of the pipe break in the high-energy line which produces the greatest 2 3.6-61 Revision 4

4.2 NuScale Power Module Piping System Parameters Table 3.6-4 lists the NuScale NPM piping along with the respective design and operating conditions. High-energy piping systems (i.e., CVCS, MSS, FWS, and DHRS) are evaluated for HELB both inside and outside the CNV. Although the DHRS condenser is manufactured from piping products, and analyzed to ASME Code, Class 2 piping rules, it is nonetheless considered a major component and not a piping system, thus breaks are not postulated. Moderate-energy piping systems (i.e., RCCWS, CFDS and CES) are exempt from HELB and are not addressed further herein. 4.3 NuScale Power Module Piping Material The high-energy piping systems are manufactured using ASME SA-312, dual-certified TP304/TP304L stainless steel, with the properties shown in Table 3.6-5, which are taken from ASME Section II, Materials. Dual-certified TP304/TP304L SS maintains the low-carbon content of the TP304L SS grade and exhibits the higher strength associated with the straight grade of TP304 SS. Thus, Table 3.6-5 uses the strength properties from the straight TP304 SS grade at design temperature of 650 degrees F shown in Table 3.6-4. Note that SA in Table 3.6-5 is calculated with a 1.0 stress range reduction factor, f. 5 References 3.6-1 Electric Power Research Institute, "An Engineering Approach for Elastic-Plastic Fracture Analysis," EPRI NP-1931, Palo Alto, CA, 1981. 3.6-2 Electric Power Research Institute, "Advances in Elastic-Plastic Fracture Analysis," EPRI NP-3607, Palo Alto, CA, 1984. 3.6-3 Electric Power Research Institute, "Elastic-Plastic Fracture Analysis of Through-Wall and Surface Flaws in Cylinders," EPRI NP-5596, Palo Alto, CA, 1988. 3.6-4 U.S. Nuclear Regulatory Commission, "Analysis of Experiments on Stainless Steel Flux Welds," NUREG/CR-4878, April 1987. 3.6-5 Not Used. 3.6-6 Electric Power Research Institute, "Crack-Opening Area Calculations for Circumferential Through-Wall Pipe Cracks," EPRI NP-5959-SR, Palo Alto, CA, 1988. 3.6-7 U.S. Nuclear Regulatory Commission, "Evaluation and Refinement of Leak-Rate Estimation Models," NUREG/CR-5128, June 1994. 2 3.6-62 Revision 4

3.6-9 Henry, R.E. and H.K. Fauske, H. K., "Two-Phase Critical Flow at Low Qualities, Part I: Experimental," Nuclear Science and Engineering. (1970): 41:79-91. 3.6-10 U.S. Nuclear Regulatory Commission, "Probabilistic Pipe Fracture Evaluations for Leak-Rate Detection Applications," NUREG/CR-6004, April 1995. 3.6-11 Not used. 3.6-12 Not used. 3.6-13 Marklund, Jan-Erik, Studsvik Energiteknik AB, Evaluation of Free Jet and Jet Impingement Tests with Hot Water and Steam, Studsvik/NR-85/54, May 21 1985. 3.6-14 Chattopadhyay, J., "Improved J and COD Estimation by GE/EPRI Method in Elastic to Fully Plastic Transition Zone," Engineering Fracture Mechanics. (2006): 73:14:1959-1979. 3.6-15 American National Standards Institute/American Nuclear Society, "Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture," ANSI/ANS-58.2-1988, LaGrange Park, IL. 3.6-16 U.S. Nuclear Regulatory Commission, Boiling Water Reactor ECCS Suction Strainer Performance Issue No. 7 - ZOI Adjustment for Air Jet Testing, BWROG Meeting, July 20, 2011, Agencywide Document Access and Management System (ADAMS) Accession No. ML11203A432. 3.6-17 U.S. Nuclear Regulatory Commission, Knowledge Base for the Effect of Debris on Pressurized Water Reactor Emergency Core Cooling Sump Performance, NUREG/CR-6808, February 2003. 3.6-18 Corradini, M., Advisory Committee on Reactor Safeguards, letter to Victor McCree, U.S. Nuclear Regulatory Commission, April 12, 2018, ADAMS Accession No. ML18102A074. 3.6-19 U.S. Nuclear Regulatory Commission, Two-Phase Jet Loads, NUREG/CR-2913, January 1983. 3.6-20 Sandia National Laboratories, "Penetration Equations," SAND-97-2426, Albuquerque, NM, October 1997. 3.6-21 NuScale Power LLC, Pipe Rupture Hazards Analysis, TR-0818-61384, Rev. 0, Corvallis, OR, September 2018. 3.6-22 Liu, J., et al., Investigation on Energetics of Ex-vessel Vapor Explosion Based on Spontaneous Nucleation Fragmentation, Journal of Nuclear Science and Technology. (2002): 39:1:31-39. 2 3.6-63 Revision 4

em Name Individual Line Names Line High- or size Moderate-Energy (NPS) Inside the Containment Vessel RCS injection 2 High RCS discharge 2 High High point vent 2 High Pressurizer spray 2 High Steam 12 & 8 High Feedwater 5&4 High S DHRS condensate return lines 1 and 2 2 High S CRDS cooling 2 Moderate Containment flooding and drain system 2 Moderate1 Outside the CNV to the NPM Disconnect Flange S RCS injection (Note 4) 4&2 High RCS discharge (Note 4) 4&2 High High point vent (Note 4) 4&2 High3 Pressurizer spray (Note 4) 4&2 High Containment evacuation 2&4 Moderate Steam 12 High Feedwater 6&5&4 High S Decay heat removal system lines 1 and 2 8&6&2 High W CRDS cooling 4&2 Moderate Containment flooding and drain system 4&2 Moderate1 In the Reactor Building (outside the NPM Disconnect Flange) Auxiliary boiler system 6 High S Containment flooding and drain system 4 High S Condensate and feedwater system 6 High S Chemical and volume control system 3 High Main steam system 12 High Module heatup system 3 High Nitrogen distribution system 2 High Process sampling system 0.75 High(2) Boron addition system 3 Moderate(1) Containment evacuation system 4 Moderate S Chilled water system 6 Moderate Demineralized water system 4 Moderate Fire protection system 16 Moderate Instrument and control air system 2 Moderate S Liquid radioactive waste system 2.5 Moderate(1) S Pool cleanup system 10 Moderate S Pool surge control system 10 Moderate WS Reactor component cooling water system 8 Moderate S Reactor pool cooling system 10 Moderate S Radioactive waste drain system 3.5 Moderate Service air system 2 Moderate Site cooling water 38 Moderate S Spent fuel pool cooling system 10 Moderate Solid radioactive waste system 3 Moderate Utility water system (5) Moderate 2 3.6-64 Revision 4

em Name Individual Line Names Line High- or size Moderate-Energy (NPS) In the Control Building S Balance-of-plant drain system 8 Moderate S Chilled water system 10 Moderate Demineralized water system 0.5 Moderate Fire protection system 16 Moderate Instrument and control air system 2 Moderate Potable water system (5) Moderate In the Radioactive Waste Building S Chilled water system 6 Moderate Demineralized water system 4 Moderate Fire protection system 12 Moderate S Gaseous radioactive waste system 2 Moderate Instrument and control air system 2 Moderate S Liquid radioactive waste system 3 Moderate Nitrogen distribution system 2 Moderate S Pool surge control system 2 Moderate S Radioactive waste drain system 3 Moderate Service air system 2 Moderate S Solid radioactive waste system 3 Moderate Outside the Control Building, Reactor Building, and Radioactive Waste Building Auxiliary boiler system 6 Moderate(1) S Balance-of-plant drain system 14 Moderate Backup power supply system (5) Moderate S Condensate and feedwater system 12 High S Chilled water system 14 Moderate Condensate polishing system 6 Moderate Circulating water system 84 Moderate Demineralized water system 6 Moderate Fire protection system 16 Moderate S Feedwater treatment system 3 High Instrument and control air system 4 Moderate S Liquid radioactive waste system 2 Moderate Main steam system 16 High Nitrogen distribution system 2 Moderate S Reactor pool surge control system 10 Moderate Potable water system (5) Moderate Process sampling system 0.75 High(2) S Radioactive waste drain system 2 Moderate Raw water system (5) Moderate Service air system 4 Moderate Site drainage system (5) Moderate S Site cooling water system 52 Moderate Turbine generator system 16 High 2 3.6-65 Revision 4

em Name Individual Line Names Line High- or size Moderate-Energy (NPS) Utility water system 36 Moderate s: ased on operating parameters that exceed 200 degrees F or 275 psig for less than 2 percent of the time the system is in peration, or that exceed 200 degrees F or 275 psig for less than 1 percent of the plant operation time. ased on the nominal diameter of the lines, breaks do not need to be postulated in PSS lines. he High point vent can be considered moderate-energy, but is conservatively evaluated as high-energy. he safe end-to-valve welds for the 2-inch CVCS lines outside the CNV are NPS4. NPS4 applies only to the single weld. ydraulic calculations have not been completed to determine system piping sizes. 2 3.6-66 Revision 4

Table 3.6-2: Postulated Break Locations Line ASME Class Postulated Break Location k locations inside containment RCS injection 1 Terminal end - RPV head Terminal end - containment boundary RCS discharge 1 Terminal end - RPV head Terminal end - containment boundary Pressurizer spray 1 Terminal end - RPV head Terminal end - containment boundary RCS high-point vent 1 Terminal end - RPV head Terminal end - containment boundary DHRS #1 2 Terminal end - containment boundary DHRS #2 2 Terminal end - containment boundary k locations outside the CNV under the bioshield None k locations in the RXB (outside the NPM bioshield) bounded, as documented in the NuScale Pipe Rupture Hazards ysis. 2 3.6-67 Revision 4

Table 3.6-3a: Not Used 2 3.6-68 Revision 4

Table 3.6-3b: Not Used 2 3.6-69 Revision 4

Table 3.6-4: NuScale Power Module Piping Systems Design and Operating Parameters cess System ASME NPS Design Operating(4) NuScale Code Size Press. Temp. Press. Temp. System) (psia) (°F) (psia) (°F) CVCS (RCS) Class 1 2 2100 650 1870(2) 625(2) CVCS NTS, CVCS) Class 3(1) 2(1) 2100 650 1870(2) 625(2) MSS (steam Class 2 8 & 12 2100 650 500 585 enerator tem, CNTS) FWS (steam Class 2 4&5 2100 650 550 300 enerator tem, CNTS) DHRS Class 2 2&6 2100 650 1400 635(3) RCCWS Class 2 2 165 200 80 121 (CRDS) RCCWS Class 2 4 1050 550 80 121 (CNTS) CFDS NTS-inside Class 2 2 165 300 85 100 CNV) CFDS TS-outside Class 2 4 1050 550 85 100 CNV) CES Class 2 4 1050 550 0.037 100 (CNTS) Notes (1) The weld between the CIV and the safe-end is NPS 4 SCH 160 and is designated as a Class 1 piping weld (2) Represents the highest normal operating pressure for the injection line and highest normal operating temperature for the RPV high point degasification line. (3) Conservatively represents the highest normal operating temperature for the steam portion (i.e., NPS 6 portion) of the DHRS. (4) The initial conditions are selected to bound system conditions for any power level, 102 percent thermal power and hot standby operation, for which the NuScale equivalent is referred to as hot shutdown. During hot shutdown, MSS pressure and temperature are approximately 300 psia and 420 degrees F, respectively, and primary pressure and temperature are approximately 1850 psia and 420 degrees F, respectively. 2 3.6-70 Revision 4

Table 3.6-5: Mechanical Properties for Piping Material Operating Room Temp Design Temp Temp ASME ystem Sy Su Sc Sy Su Sm Sh SA E Sy (ksi) Class (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (106 psi) CVCS 1 16.2 NA NA 18.2 RCS) CVCS CNTS, 3 18.2 30 75 20.0 18.0 63.4 25.1 VCS) NA 16.2 29.05 FWS 2 22.4 MSS 2 18.6 DHRS 2 18.1 2 3.6-71 Revision 4

Process System 2.4Sm 2.25Sm 1.8Sy 1.2Sm CVCS (RCS) 38.88 36.45 32.40 19.44 2 3.6-72 Revision 4

Process System 0.8(1.8Sh+SA) 2.25Sh 1.8Sy 0.4(1.8Sh+SA) VCS (CNTS, CVCS) FWS 46.57 36.45 32.40 23.28 MSS DHRS 2 3.6-73 Revision 4

2 3.6-74 Revision 4 Protection against Dynamic Effects Associated with Postulated Rupture cale Final Safety Analysis Report of Piping

Figure 3.6-2: Not Used 2 3.6-76 Revision 4

2 3.6-77 Revision 4 2 3.6-78 Revision 4 2 3.6-79 Revision 4 2 3.6-80 Revision 4 2 3.6-81 Revision 4 2 3.6-82 Revision 4 2 3.6-83 Revision 4 2 3.6-84 Revision 4 2 3.6-85 Revision 4 2 3.6-86 Revision 4 2 3.6-87 Revision 4 2 3.6-88 Revision 4 2 3.6-89 Revision 4 Protection against Dynamic Effects Associated with Postulated Rupture cale Final Safety Analysis Report of Piping

Protection against Dynamic Effects Associated with Postulated Rupture cale Final Safety Analysis Report of Piping

Start LBB for Candidate Piping Screening of Potential Degradation Mechanisms Geometry and Material Properties Low Normal Stress Intermediate Normal Stresses High Normal Stress Crack Opening Areas at Each Leak Rate = Leak Detection Normal Stress Level Capability*10 Leakage Crack Length CL Maximum Stress by Limit Load Analysis for 2CL Crack Length by Thermal-Hydraulic Model Normal Loads Normal and Plot SBAC with Low, Intermediate Maximum and High Normal Stresses and Their Stress Point P* Corresponding Maximum Stresses Maximum Loads No Is Point P* Yes below SBAC? HELB Not Qualified for LBB Qualified for LBB SRP 3.6.2 2 3.6-92 Revision 4

2 3.6-93 Revision 4 2 3.6-94 Revision 4 2 3.6-95 Revision 4 2 3.6-96 Revision 4 Straight Pipe Base Metal

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