Text

Part 02 - Final Safety Analysis Report (Rev. 1) - Part 02 - Tier 02 - Chapter 03 - Design of Structures, Systems, Components and Equipment - Sections 03.01 - 03.06
	ML18086A153
Person / Time
Site:	NuScale
Issue date:	03/15/2018
From:	Bergman T; NuScale
To:	; Office of New Reactors
	Cranston G
References
	NUSCALESMRDC, NUSCALESMRDC.SUBMISSION.4, NUSCALEPART02.NP, NUSCALEPART02.NP.1
	Download: ML18086A153 (232)
	v • d • e

NuScale Standard Plant Design Certification Application Chapter Three Design of Structures, Systems, Components and Equipment PART 2 - TIER 2 Revision 1 March 2018

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APTER 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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-15 3.1.4 Fluid Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-22 3.1.5 Reactor Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-36 3.1.6 Fuel and Radioactivity Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-42 3.1.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.1-46 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 . . . . . . . . . . . . . . . . 3.4-6 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. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-6 3.5.3 Barrier Design Procedures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-7 3.5.4 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-12 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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-6 2 i Revision 1

3.6.3 Leak-Before-Break Evaluation Procedures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-16 3.6.4 High Energy Line Break Evaluation (Non-LBB) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-41 3.6.5 Integral Jet Impingement Shield and Pipe Whip Restraint . . . . . . . . . . . . . . . . . . . 3.6-44 3.6.6 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-50 3.7 Seismic Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-1 3.7.1 Seismic Design Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-2 3.7.2 Seismic System Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-104 3.7.3 Seismic Subsystem Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-349 3.7.4 Seismic Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-375 3.7.5 Computer Programs Used in Section 3.7 Seismic Design . . . . . . . . . . . . . . . . . . . 3.7-378 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 . . . 3.8-38 3.8.4 Other Seismic Category I Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-39 3.8.5 Foundations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-111 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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-34 3.9.5 Reactor Vessel Internals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-43 3.9.6 Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic Restraints. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-48 3.9.7 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-60 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 2 ii Revision 1

3.10.5 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.10-8 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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-11 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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-1 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 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-15 3.12.6 Piping Support Design Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-24 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-30 3B.4 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-34 2 iii Revision 1

Appendix 3C Methodology for Environmental Qualification of Electrical and Mechanical Equipment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-1 3C.1 Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-1 3C.2 Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-1 3C.3 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-1 3C.4 Qualification Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-2 3C.5 Design Specifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3C-8 3C.6 Qualification Methods . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-12 3C.7 Equipment Qualification Maintenance Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-15 3C.8 Documentation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-16 3C.9 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-16 2 iv Revision 1

le 3.2-1: Classification of Structures, Systems, and Components . . . . . . . . . . . . . . . . . . . . . . . . . 3.2-8 le 3.4-1: Flooding Sources in the Reactor Building and Control Building . . . . . . . . . . . . . . . . 3.4-10 le 3.4-2: Flood Levels for Rooms Containing Systems, Structures, and Components Subject to Flood Protection (Without Mitigation) . . . . . . . . . . . . . . . . 3.4-11 le 3.5-1: Concrete Thickness to Preclude Missile Penetration, Perforation, or Scabbing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-13 le 3.6-1: High- and Moderate-Energy Fluid System Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-52 le 3.6-2: Postulated Break Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-55 le 3.6-3a: Summary of Main Steam Line Bounding Analysis Curves . . . . . . . . . . . . . . . . . . . . . . 3.6-56 le 3.6-3b: Summary of Feedwater System Line Bounding Analysis Curves . . . . . . . . . . . . . . . . 3.6-57 le 3.6-4: NuScale Power Module Piping Systems Design and Operating Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-58 le 3.6-5: Mechanical Properties for Piping Material . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-59 le 3.6-6: Allowable Stresses for Class 1 Piping (ksi) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-60 le 3.6-7: Allowable Stresses for Class 2 & 3 Piping (ksi). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-61 le 3.6-8: Jet loads and Maximum Bending Moments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-62 le 3.7.1-1: Certified Seismic Design Response Spectra Control Points at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-19 le 3.7.1-2: Certified Seismic Design Response Spectra - High Frequency Control Points at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-20 le 3.7.1-3: Cross-Correlation Coefficients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-21 le 3.7.1-4: Duration of Time Histories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-22 le 3.7.1-5: Comparison of Response Spectra to CSDRS and CSDRS-HF . . . . . . . . . . . . . . . . . . . . 3.7-23 le 3.7.1-6: Generic Damping Values for Dynamic Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-24 le 3.7.1-7: Effective Stiffness of Reinforced Concrete Members . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-26 le 3.7.1-8: Soil Shear Modulus Degradation and Strain-Dependent Soil Damping (0-120 ft) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-27 le 3.7.1-9: Soil Shear Modulus Degradation and Strain-Dependent Soil Damping (120 ft-1000 ft) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-28 le 3.7.1-10: Strain-Dependent Soil Shear Moduli and Soil Damping Ratios for Gravel and Rock. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-29 le 3.7.1-11: Soft Soil [Type 11] Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-30 le 3.7.1-12: Firm Soil/Soft Rock [Type 8] Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-31 le 3.7.1-13: Rock [Type 7] Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-32 le 3.7.1-14: Hard Rock [Type 9] Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-33 2 v Revision 1

le 3.7.1-15: Average Strain-Compatible Properties for CSDRS for Rock [Type 7]. . . . . . . . . . . . . 3.7-34 le 3.7.1-16: Average Strain-Compatible Properties for CSDRS for Soft Soil [Type 11] . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-36 le 3.7.1-17: Average Strain-Compatible Properties for CSDRS for Firm Soil/Soft Rock [Type 8] . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-38 le 3.7.1-18: Strain-Compatible Properties for CSDRS-HF for Rock [Type 7] . . . . . . . . . . . . . . . . . . 3.7-40 le 3.7.1-19: Strain-Compatible Properties for CSDRS-HF for Hard Rock [Type 9]. . . . . . . . . . . . . 3.7-42 le 3.7.1-20: Wave Passing Frequencies. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-44 le 3.7.1-21: Shear Wave Fundamental Frequencies of Soil Columns above RXB Foundation Bottom Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-45 le 3.7.2-1: Summary of Reactor Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-140 le 3.7.2-2: Average Hydrodynamic Pressure from ANSYS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-141 le 3.7.2-3: Equivalent Average Static Pressure from SASSI2010 . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-142 le 3.7.2-4: Summary of Average Pressures and Equivalent Static Pressure for SASSI2010 Soil Type 7. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-143 le 3.7.2-5: Summary of Average Pressures and Equivalent Static Pressure for SASSI2010 Soil Type 8. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-144 le 3.7.2-6: Summary of Average Pressures and Equivalent Static Pressure for SASSI2010 Soil Type 11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-145 le 3.7.2-7: Comparison of Pressures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-146 le 3.7.2-8: Final Surface Pressure Adjustment in SAP2000 Model Due to FSI Effects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-147 le 3.7.2-9: Summary of Control Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-148 le 3.7.2-10: Summary of Reactor Building Fixed-Base Modal Frequency Comparison . . . . . . 3.7-149 le 3.7.2-11: Summary of Control Building Fixed-Base Model Frequency Comparison . . . . . . 3.7-150 le 3.7.2-12: Summary of Triple Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-151 le 3.7.2-13: Dimensions and Weights of the Three Buildings. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-152 le 3.7.2-14: Frequencies and Modal Mass Ratios for the Reactor Building Cracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-153 le 3.7.2-15: Frequencies and Modal Mass Ratios for the Reactor Building Uncracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-164 le 3.7.2-16: Frequencies and Modal Mass Ratios for the Control Building Cracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-175 le 3.7.2-17: Frequencies and Modal Mass Ratios for the Control Building Uncracked Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-180 2 vi Revision 1

le 3.7.2-18: Frequencies Used in Transfer Function Calculation for Standalone Reactor Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-185 le 3.7.2-19: Frequencies Used in Transfer Function Calculation for RXB from Triple Building Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-188 le 3.7.2-20: Frequencies Used in Transfer Function Calculation for Standalone CRB Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-191 le 3.7.2-21: Frequencies Used in Transfer Function Calculation for CRB with Triple Building CRB Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-195 le 3.7.2-22: Methodology for Combining SASSI2010 Results. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-199 le 3.7.2-23: Example Averaging and Bounding Forces and Moments in a Shell Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-201 le 3.7.2-24: Example Averaging and Bounding Forces and Moments in a Beam Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-202 le 3.7.2-25: Example Averaging and Bounding Forces and Moments in a Solid Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-204 le 3.7.2-26: Selected Reactor Building Locations for Relative Displacement . . . . . . . . . . . . . . . 3.7-205 le 3.7.2-27: Selected Control Building Locations for Relative Displacement Calculation. . . . 3.7-206 le 3.7.2-28: Relative Displacement at Selected Locations on Reactor Building . . . . . . . . . . . . . 3.7-207 le 3.7.2-29: Relative Displacement at Selected Locations on Control Building . . . . . . . . . . . . . 3.7-208 le 3.7.2-30: Comparison of Maximum Lug and Skirt Reactions using Soil Type 7 (CSDRS). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-209 le 3.7.2-31: Comparison of Maximum Lug and Skirt Reactions using Soil Type 9 (CSDRS-HF) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-210 le 3.7.2-32: Max Forces and Moments at wall locations using Soil Type 7, CSDRS Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-211 le 3.7.2-33: Definition of Seismic Analysis Identification Codes . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-212 le 3.7.2-34: SSC Seismic Analysis Identification Code Assignments. . . . . . . . . . . . . . . . . . . . . . . . 3.7-213 le 3.7.2-35: Analysis Model Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-214 le 3.7.2-36: SASSI2010 3D Equivalent Stick Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-218 le 3.7.2-37: ANSYS 3D Finite Element Beam Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-219 le 3.7.2-38: Major Mode Comparisons Between Simplified NuScale Power Module Beam Model and 3-D Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-220 le 3.7.3-1: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-357 le 3.7.3-2: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-358 le 3.7.3-3: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-359 le 3.7.3-4: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-360 2 vii Revision 1

le 3.7.3-5: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-361 le 3.7.3-6: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-362 le 3.7.3-7: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-363 le 3.7.3-8: Bioshield Nominal. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-364 le 3.7.3-9: Bioshield Concrete and Reinforcement Design Properties . . . . . . . . . . . . . . . . . . . . 3.7-365 le 3.7.3-10: Moment and Shear Capacity of Horizontal Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-366 le 3.7.3-11: Bioshield Slab Self-Weight. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-367 le 3.7.3-12: Bioshield Face Plate Self-Weight . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-368 le 3.7.3-13: Horizontal Bioshield Accelerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-369 le 3.7.3-14: Summary of Bioshield Demand to Capacity Ratios. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-370 le 3.8.2-1: Design and Operating Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-25 le 3.8.2-2: Load Combinations for Containment Vessel and Support ASME Code Stress Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-26 le 3.8.2-3: Load Combinations for Containment Vessel Bolt ASME Code Stress Analysis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-27 le 3.8.2-4: Key Assumptions for CNV Ultimate Pressure Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-28 le 3.8.4-1: Concrete Design Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-70 le 3.8.4-2: Steel Design Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-71 le 3.8.4-3: Summary of Reactor Building Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-72 le 3.8.4-4: Summary of Control Building Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-73 le 3.8.4-5: Hydrodynamic Weight . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-74 le 3.8.4-6: Reactor Building SAP2000 Joints and Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-75 le 3.8.4-7: Reactor Building SAP2000 Mass Sources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-76 le 3.8.4-8: Control Building SAP2000 Joints and Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-77 le 3.8.4-9: Control Building SAP2000 Mass Sources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-78 le 3.8.4-10: Material Properties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-79 le 3.8.4-11: Additional Dynamic Analyses. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-80 le 3.8.4-12: Seismic Categories and Design Codes. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-81 le 3.8.4-13: Total Weight in Kips, SAP2000 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-82 le 3.8.4-14: Total Weight in Kips, SASSI2010 Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-83 le 3.8.5-1: RXB Stability Evaluation Input Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-132 le 3.8.5-2: Reactor Building Static Effective Soil Force . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-133 le 3.8.5-3: Seismic Base Reactions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-134 2 viii Revision 1

le 3.8.5-4: Seismic Vertical RXB Base Reactions and Dead Weight. . . . . . . . . . . . . . . . . . . . . . . . 3.8-136 le 3.8.5-5: Factors of safety - RXB Stability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-137 le 3.8.5-6: RXB ANSYS Model Summary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-138 le 3.8.5-7: Overturning Forces and Overturning Arms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-139 le 3.8.5-8: Settlement values for the RXB, CRB and RWB . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-140 le 3.8.5-9: CRB Stability Input Evaluation Parameters. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-141 le 3.8.5-10: CRB Total Static Lateral Soil Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-142 le 3.8.5-11: CRB SAP2000, SASSI2010, and ANSYS Model Summary . . . . . . . . . . . . . . . . . . . . . . . 3.8-143 le 3.8.5-12: Reactor Building Sliding Displacements for Soil Type 7, 8 and 11 (Dead Weight + Buoyancy) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-144 le 3.8.5-13: Control Building Sliding and Uplift Displacements for Soil Type 7 and 11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-145 le 3.8.5-14: Average Soil Bearing Pressures (Toe Pressures) along Edges of RXB Basemat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-146 le 3.8.5-15: Seismic Vertical CRB Base Reactions and Dead Weight. . . . . . . . . . . . . . . . . . . . . . . . 3.8-147 le 3.8.5-16: Average Soil Bearing Pressures (Toe Pressures) along Edges of CRB Basemat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-148 le 3.8.5-17: Reactor Building SAP2000 Basemat Model Summary . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-149 le 3.9-1: Summary of Design Transients . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-62 le 3.9-2: Pressure, Mechanical, and Thermal Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-63 le 3.9-3: Required Load Combinations for Reactor Pressure Vessel American Society of Mechanical Engineers Stress Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-64 le 3.9-4: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-65 le 3.9-5: Required Load Combinations for Reactor Vessel Internals American Society of Mechanical Engineers Stress Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-66 le 3.9-6: Required Load Combinations for Control Rod Drive Mechanism American Society of Mechanical Engineers Stress Analysis . . . . . . . . . . . . . . . . . . . . . 3.9-67 le 3.9-7: Load Combinations for Decay Heat Removal System Condenser . . . . . . . . . . . . . . . 3.9-68 le 3.9-8: Load Combinations for NuScale Power Module Top Support Structure. . . . . . . . . 3.9-69 le 3.9-9: Loading Combinations for Decay Heat Removal System Actuation Valves . . . . . . 3.9-70 le 3.9-10: Loads and Load Combinations for Reactor Safety Valves. . . . . . . . . . . . . . . . . . . . . . . 3.9-71 le 3.9-11: Load Combinations for Emergency Core Cooling System Valves . . . . . . . . . . . . . . . 3.9-72 le 3.9-12: Required Loads and Load Combinations for Secondary System Containment Isolation Valves. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-73 2 ix Revision 1

le 3.9-13: Required American Society of Mechanical Engineers Code Loads and Load Combinations for Primary System Containment Isolation Valves. . . . . . . . . . 3.9-74 le 3.9-14: Loads and Load Combinations for Thermal Relief Valves. . . . . . . . . . . . . . . . . . . . . . . 3.9-75 le 3.9-15: Active Valve List. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-76 le 3.9-16: Valve Inservice Test Requirements per ASME OM Code . . . . . . . . . . . . . . . . . . . . . . . . 3.9-78 le 3.9-17: Valve Augmented Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-87 le 3.9-18: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-90 le 3.9-19: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-91 le 3.9-20: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-92 le 3.9-21: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-93 le 3.9-22: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-94 le 3.9-23: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-95 le 3.9-24: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-96 le 3.9-25: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-97 le 3.9-26: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-98 le 3.11-1: List of Environmentally Qualified Electrical/I&C and Mechanical Equipment Located in Harsh Environments. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.11-13 le 3.11-2: Environmental Qualification Zones - Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . 3.11-26 le 3.12-1: Required Load Combinations for Class 1 Piping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-32 le 3.12-2: Required Load Combinations for Class 2 & 3 Piping. . . . . . . . . . . . . . . . . . . . . . . . . . . 3.12-34 le 3.12-3: Required Load Combinations for Class 1, 2, & 3 Supports . . . . . . . . . . . . . . . . . . . . . 3.12-35 le 3.13-1: ASME BPV Code Section III Criteria for Selection and Testing of Bolted Materials. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.13-5 le 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 le 3B-1: Identification of SAP2000 and SASSI2010 Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-36 le 3B-2: Summary of D/C Ratios for Reactor Building Wall at Grid Line 1 . . . . . . . . . . . . . . . . .3B-37 le 3B-3: Summary of D/C Ratios for Reactor Building Wall at Grid Line 3 . . . . . . . . . . . . . . . . .3B-39 le 3B-4: Element Averaging of Horizontal Reinforcement Exceedance for Reactor Building Wall at Grid Line 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-40 le 3B-5: Element Averaging of Horizontal Membrane Compression Stress for Reactor Building Wall at Grid Line 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-41 le 3B-6: Element Averaging of Vertical Reinforcement Exceedance for Reactor Building Wall at Grid Line 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-42 2 x Revision 1

le 3B-7: Summary of D/C Ratios for Reactor Building Wall at Grid Line 3 After Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-43 le 3B-8: Summary of D/C Ratios for Reactor Building Wall at Grid Line 4 . . . . . . . . . . . . . . . . .3B-44 le 3B-9: Element Averaging of Reinforcement Exceedance for Reactor Building Wall at Grid Line 4. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-45 le 3B-10: Summary of D/C Ratios for RXB Wall at Grid Line 4 After Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-46 le 3B-11: Summary of D/C Ratios for RXB Wall at Grid Line 6. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-47 le 3B-12: Element Averaging of Horizontal Reinforcement Exceedance for RXB Wall at Grid Line 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-49 le 3B-13: Summary of D/C Ratios for Reactor Building Wall at Grid Line 6 after Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-50 le 3B-14: Summary of D/C Ratios for Reactor Building Wall at Grid Line E . . . . . . . . . . . . . . . . .3B-52 le 3B-15: Summary of D/C Ratios for Reactor Building Slab at EL. 100-0. . . . . . . . . . . . . . . . . .3B-55 le 3B-16: Element Averaging of XZ Plane Shear Exceedance for Reactor Building Slab at EL. 100-0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-57 le 3B-17: Summary of D/C Ratios for Reactor Building Slab at EL. 100-0 After Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-58 le 3B-18: Summary of D/C Ratios for RXB Roof Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-60 le 3B-19: Summary of D/C Ratios for Reactor Building Pilasters on Grid Line A Wall. . . . . . . .3B-62 le 3B-20: Summary of D/C Ratios for Reactor Building Beams on EL. 75'-0" Slab . . . . . . . . . . .3B-64 le 3B-21: Summary of D/C Ratios for Reactor Building Buttress at Grid Line 1 on EL. 126'-0" Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-66 le 3B-22: Summary of D/C Ratios for West Wing Wall at Grid Line 4 . . . . . . . . . . . . . . . . . . . . . . 3.B-67 le 3B-23: Summary of D/C Ratios for Reactor Building Pool Wall at Grid Line B . . . . . . . . . . . 3.B-68 le 3B-24: Element Averaging of YZ Plane Shear Exceedance for Reactor Building Pool Wall at Grid Line B. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.B-70 le 3B-25: Summary of D/C Ratios for Reactor Building Pool Wall at Grid Line B After Averaging Affected Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-71 le 3B-26: NuScale Power Module Lug Support Model Cut Section Forces and Moments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.B-73 le 3B-27: SASSI Maximum Lug Reactions for RXB Cracked Model using Soil Type 7 (CSDRS) and Soil Type 9 (CSDRS-HF) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-74 le 3B-28: Summary of D/C Ratios for Control Building Wall at Grid Line 3 . . . . . . . . . . . . . . . . .3B-75 le 3B-29: Summary of D/C Ratios for Control Building Wall at Grid Line 4 . . . . . . . . . . . . . . . . .3B-76 2 xi Revision 1

le 3B-30: Control Building Wall at Grid Line 4 - Shell Element 786 with added Shear Reinforcement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-77 le 3B-31: Summary of D/C Ratios for Control Building Wall at Grid Line 4 After Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-78 le 3B-32: Summary of D/C Ratios for Control Building Wall at Grid Line A . . . . . . . . . . . . . . . . .3B-80 le 3B-33: Element Averaging of IP Shear Exceedance of Control Building Wall at Grid Line A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-81 le 3B-34: Moment and Shear Capacity: 5 Foot Thick Control Building Basemat Foundation (Type 1). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-82 le 3B-35: Moment and Shear Capacity: 5 Foot Thick Control Building Basemat Foundation (Type 2). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-83 le 3B-36: Magnitudes of Bounding Demand Forces and Moments for Perimeter of Main Control Building Basemat Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-84 le 3B-37: Magnitudes of Bounding Demand Forces and Moments for Interior of Main Control Building Basemat Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-85 le 3B-38: Magnitudes of Bounding Demand Forces and Moments for Control Building Basemat of Control Building Tunnel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-86 le 3B-39: Design Check Control Building Basemat Foundation of Perimeter of the Main Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-87 le 3B-40: Design Check Control Building Basemat Foundation of Interior of the Main Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-88 le 3B-41: Design Check for Control Building Basemat Foundation for the Control Building Tunnel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-89 le 3B-42: Summary of D/C Ratios for Control Building Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . .3B-90 le 3B-43: Element Averaging of East-West Reinforcement Exceedance - Control Building Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-91 le 3B-44: Element Averaging of XZ Plane Shear Exceedance - Control Building Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-92 le 3B-45: Element Averaging of YZ Plane Shear Exceedance - Control Building Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-93 le 3B-46: Summary of D/C Ratios for Control Building Slab at EL. 100'-0" After Averaging Affected Elements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-94 le 3B-47: Element Averaging of Shear Friction Exceedance for Control Building Slab at EL. 100-0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-95 le 3B-48: Summary of D/C Ratios for Control Building Pilasters on Grid Line 1 Wall . . . . . . . .3B-96 le 3B-49: Summary of D/C Ratios for Control Building T-Beams on EL. 120'-0" Slab . . . . . . . .3B-97 le 3B-50: Element Averaging of IP Shear Exceedance of Reactor Building Wall at Grid Line 3. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-98 2 xii Revision 1

le 3B-51: Element Averaging of Shear Friction Exceedance of Reactor Building Wall at Grid Line 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3B-99 le 3B-52: Analysis Cases for NuScale Power Modules. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-100 le 3B-53: Strength Reduction Factors for Reinforced Concrete Design . . . . . . . . . . . . . . . . . . 3B-101 le 3B-54: RXB Critical Sections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-102 le 3B-55: CRB Critical Sections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-103 le 3C-1: Environmental Qualification Zones - Reactor Building . . . . . . . . . . . . . . . . . . . . . . . . . .3C-18 le 3C-2: Designated Harsh Environment Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-19 le 3C-3: Designated Mild Environment Areas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-20 le 3C-4: Equipment Post-Accident Operating Times . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-21 le 3C-5: EQ Program Margin Requirements. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-23 le 3C-6: Normal Operating Environmental Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-24 le 3C-7: Design Basis Event Environmental Conditions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-25 le 3C-8: Accident EQ Radiation Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-26 2 xiii Revision 1

re 3.6-1: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-63 re 3.6-2: Main Steam Line 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-64 re 3.6-3: Main Steam Line 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-65 re 3.6-4: Feedwater Line 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-66 re 3.6-5: Feedwater Line 2. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-67 re 3.6-6: Chemical and Volume Control System - Reactor Coolant System Injection Line Postulated Break Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-68 re 3.6-7: Chemical and Volume Control System - Reactor Coolant System Discharge Line Postulated Break Locations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-69 re 3.6-8: Chemical and Volume Control System - Pressurizer Spray Line Postulated Break Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-70 re 3.6-9: Chemical and Volume Control System - High Point Vent Postulated Break Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-71 re 3.6-10: Decay Heat Removal System Line 1 Postulated Break Locations. . . . . . . . . . . . . . . . 3.6-72 re 3.6-11: Decay Heat Removal System Line 2 Postulated Break Locations. . . . . . . . . . . . . . . . 3.6-73 re 3.6-12: Containment System Chemical and Volume Control Discharge and Injection Line Postulated Break Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-74 re 3.6-13: Chemical and Volume Control System Postulated Break Locations . . . . . . . . . . . . . 3.6-75 re 3.6-14: Feedwater Line Postulated Break Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-76 re 3.6-15: Main Steam Line Postulated Break Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-77 re 3.6-16: Postulated High-Energy Main Steam System Pipe Routing Beyond the NuScale Power Module (COL Applicant Scope) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-78 re 3.6-17: Postulated High-Energy Feedwater System Pipe Routing Beyond the NuScale Power Module (COL Applicant Scope) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-79 re 3.6-18: Flow Chart for Piping Leak-Before-Break Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-80 re 3.6-19: Illustration of Pipe with a Circumferential Through-Wall Crack . . . . . . . . . . . . . . . . . 3.6-81 re 3.6-20: Henry-Fauske's Model of Two-Phase Flow. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-82 re 3.6-21: Local and Global Surface Roughness and Turns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-83 re 3.6-22: Crack Opening Displacement-Dependent Effective Crack Morphology . . . . . . . . . 3.6-84 re 3.6-23: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Straight Pipe Base Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-85 re 3.6-24: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Pipe-to-Pipe Weld. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-86 re 3.6-25: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Pipe-to-Safe-End Weld . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-87 2 xiv Revision 1

re 3.6-26: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 12 Straight Pipe Base Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-88 re 3.6-27: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 12 Pipe-to-Safe-End Weld . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-89 re 3.6-28: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Elbow Base Metal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-90 re 3.6-29: Smooth Bounding Analysis Curve for Nominal Pipe Size 4 Feedwater System Line Base Metal. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-91 re 3.6-30: Smooth Bounding Analysis Curve for Nominal Pipe Size 4 Feedwater System Line Welds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-92 re 3.6-31: Smooth Bounding Analysis Curve for Nominal Pipe Size 5 Feedwater System Line Base Metal. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-93 re 3.6-32: Smooth Bounding Analysis Curve for Nominal Pipe Size 5 Feedwater System Line Welds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-94 re 3.6-33: Typical Integral Jet Impingement Shield and Pipe Whip Restraint . . . . . . . . . . . . . . 3.6-95 re 3.6-34: Cutaway View of Integral Jet Impingement Shield and Pipe Whip Restraint . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-96 re 3.6-35: Disk-Type Jet from Circumferential Pipe Rupture at a Weld . . . . . . . . . . . . . . . . . . . . 3.6-97 re 3.6-36: RCS pipe break total pressure drop along discharge centerline . . . . . . . . . . . . . . . . 3.6-98 re 3.6-37: RCS pipe break total pressure graph at 5, 10, 15, and 20 inches radially from ISR. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-99 re 3.6-38: DHRS high temperature pipe break total pressure graph along discharge centerline . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-100 re 3.6-39: DHRS low temperature pipe break total pressure drop along discharge centerline. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-101 re 3.6-40: DHRS low temperature pipe break total pressure graph at 5, 10, 15, and 20 inches radially from ISR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-102 re 3.7.1-1: NuScale Horizontal CSDRS at 5 Percent Damping. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-46 re 3.7.1-2: NuScale Vertical CSDRS at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-47 re 3.7.1-3: NuScale Horizontal CSDRS-HF at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-48 re 3.7.1-4: NuScale Vertical CSDRS-HF at 5 Percent Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-49 re 3.7.1-5a: Original Time Histories for Yermo East-West. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-50 re 3.7.1-5b: CSDRS Compatible Time Histories for Yermo East-West. . . . . . . . . . . . . . . . . . . . . . . . 3.7-51 re 3.7.1-5c: Original Time Histories for Yermo North-South . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-52 re 3.7.1-5d: CSDRS Compatible Time Histories for Yermo North-South . . . . . . . . . . . . . . . . . . . . . 3.7-53 re 3.7.1-5e: Original Time Histories for Yermo Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-54 2 xv Revision 1

re 3.7.1-5f: CSDRS Compatible Time Histories for Yermo Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-55 re 3.7.1-6a: Original Time Histories for Capitola East-West . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-56 re 3.7.1-6b: CSDRS Compatible Time Histories for Capitola East-West . . . . . . . . . . . . . . . . . . . . . . 3.7-57 re 3.7.1-6c: Original Time Histories for Capitola North-South . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-58 re 3.7.1-6d: CSDRS Compatible Time Histories for Capitola North-South . . . . . . . . . . . . . . . . . . . 3.7-59 re 3.7.1-6e: Original Time Histories for Capitola Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-60 re 3.7.1-6f: CSDRS Compatible Time Histories for Capitola Vertical . . . . . . . . . . . . . . . . . . . . . . . . 3.7-61 re 3.7.1-7a: Original Time Histories for Chi-Chi East-West . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-62 re 3.7.1-7b: CSDRS Compatible Time Histories for Chi-Chi East-West . . . . . . . . . . . . . . . . . . . . . . . 3.7-63 re 3.7.1-7c: Original Time Histories for Chi-Chi North-South . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-64 re 3.7.1-7d: CSDRS Compatible Time Histories for Chi-Chi North-South . . . . . . . . . . . . . . . . . . . . 3.7-65 re 3.7.1-7e: Original Time Histories for Chi-Chi Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-66 re 3.7.1-7f: CSDRS Compatible Time Histories for Chi-Chi Vertical . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-67 re 3.7.1-8a: Original Time Histories for Izmit East-West . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-68 re 3.7.1-8b: CSDRS Compatible Time Histories for Izmit East-West . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-69 re 3.7.1-8c: Original Time Histories for Izmit North-South . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-70 re 3.7.1-8d: CSDRS Compatible Time Histories for Izmit North-South. . . . . . . . . . . . . . . . . . . . . . . 3.7-71 re 3.7.1-8e: Original Time Histories for Izmit Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-72 re 3.7.1-8f: CSDRS Compatible Time Histories for Izmit Vertical. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-73 re 3.7.1-9a: Original Time Histories for El Centro East-West . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-74 re 3.7.1-9b: CSDRS Compatible Time Histories for El Centro East-West . . . . . . . . . . . . . . . . . . . . . 3.7-75 re 3.7.1-9c: Original Time Histories for El Centro North South. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-76 re 3.7.1-9d: CSDRS Compatible Time Histories for El Centro North-South. . . . . . . . . . . . . . . . . . . 3.7-77 re 3.7.1-9e: Original Time Histories for El Centro Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-78 re 3.7.1-9f: CSDRS Compatible Time Histories for El Centro Vertical. . . . . . . . . . . . . . . . . . . . . . . . 3.7-79 re 3.7.1-10a: Original Time Histories for Lucerne East-West . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-80 re 3.7.1-10b: CSDRS-HF Compatible Time Histories for Lucerne East-West . . . . . . . . . . . . . . . . . . . 3.7-81 re 3.7.1-10c: Original Time Histories for Lucerne North-South . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-82 re 3.7.1-10d: CSDRS-HF Compatible Time Histories for Lucerne North-South . . . . . . . . . . . . . . . . 3.7-83 re 3.7.1-10e: Original Time Histories for Lucerne Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-84 re 3.7.1-10f: CSDRS-HF Compatible Time Histories for Lucerne Vertical . . . . . . . . . . . . . . . . . . . . . 3.7-85 2 xvi Revision 1

re 3.7.1-11: Normalized Arias Intensity Curve of North-South Component of Izmit Time History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-86 re 3.7.1-12a: Average Response Spectra East-West . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-87 re 3.7.1-12b: Average Response Spectra North-South . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-88 re 3.7.1-12c: Average Response Spectra Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-89 re 3.7.1-13a: Power Spectral Density Curves CSDRS Compatible Time Histories. . . . . . . . . . . . . . 3.7-90 re 3.7.1-13b: Power Spectral Density Curves CSDRS-HF Compatible Time Histories . . . . . . . . . . 3.7-91 re 3.7.1-14: Soil Shear Modulus Degradation Curves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-92 re 3.7.1-15: Strain Dependent Soil Damping Curves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-93 re 3.7.1-16: Shear Wave Velocities for All Soil Types . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-94 re 3.7.1-17: Layered Soil Model Used for NuScale Power Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-95 re 3.7.1-18: Density for All Soil Types . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-96 re 3.7.1-19: Average Strain Compatible Vs Profiles for CSDRS Compatible Inputs . . . . . . . . . . . 3.7-97 re 3.7.1-20: Strain Compatible Vs Profiles for CSDRS-HF Compatible Input . . . . . . . . . . . . . . . . . 3.7-98 re 3.7.1-21: Strain Compatible Damping for Soil Type 7 for CSDRS Compatible Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-99 re 3.7.1-22: Strain Compatible Damping for Soil Type 8 for CSDRS Compatible Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-100 re 3.7.1-23: Strain Compatible Damping for Soil Type 11 for CSDRS Compatible Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-101 re 3.7.1-24: Comparison of Average Strain Compatible Damping for CSDRS Compatible Inputs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-102 re 3.7.1-25: Comparison of Strain Compatible Damping for CSDRS-HF Compatible Input . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-103 re 3.7.2-1: Control Building, Reactor Building, and Radioactive Waste Building in Soil (Looking Northeast). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-221 re 3.7.2-2: Section View of Control Building, Reactor Building, and Radioactive Waste Building in Soil (Looking Northeast) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-222 re 3.7.2-3: Global Origin of Building Models . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-223 re 3.7.2-4: Reactor Building Model Showing Global X, Y, and Z Axes at Origin . . . . . . . . . . . . 3.7-224 re 3.7.2-5: Location at Northeast Corner on Top of Basemat used for 7P versus 9P Comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-225 re 3.7.2-6: Location at NPM 1 East Wing Wall at Lug Support used for 7P versus 9P Comparison. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-226 re 3.7.2-7: Location at Center of Roof Slab used for 7P versus 9P Comparison . . . . . . . . . . . . 3.7-227 2 xvii Revision 1

re 3.7.2-8: 7P Versus 9P Comparison at Northeast Corner on Top of Basemat . . . . . . . . . . . . 3.7-228 re 3.7.2-9: 7P Versus 9P Comparison at NPM 1 East Wing Wall at Lug Support. . . . . . . . . . . . 3.7-229 re 3.7.2-10: 7P Versus 9P Comparison at Center of Roof Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-230 re 3.7.2-11: Reactor Building in Ground (Looking Northeast) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-231 re 3.7.2-12: Quarter View of Reactor Building in Ground . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-232 re 3.7.2-13: Longitudinal View of Half of Reactor Building in Ground. . . . . . . . . . . . . . . . . . . . . . 3.7-233 re 3.7.2-14: Transverse View of Half of Reactor Building in Ground (Looking Northeast) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-234 re 3.7.2-15: Reactor Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-235 re 3.7.2-16: Reactor Building SASSI2010 Model without Hidden Lines. . . . . . . . . . . . . . . . . . . . . 3.7-236 re 3.7.2-17: Reactor Building SASSI2010 Backfill Soil Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-237 re 3.7.2-18: Reactor Building SASSI2010 Model without Backfill. . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-238 re 3.7.2-19: Reactor Building SASSI2010 Excavated Soil Model without Hidden Lines . . . . . . 3.7-239 re 3.7.2-20: Half of Reactor Building SASSI2010 Model without Hidden Lines. . . . . . . . . . . . . . 3.7-240 re 3.7.2-21: Reactor Building Beam Elements of SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . 3.7-241 re 3.7.2-22: NuScale Power Module Lug Restraint (in Green) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-242 re 3.7.2-23: Top View of NuScale Power Module Lug Restraint and Support Walls . . . . . . . . . 3.7-243 re 3.7.2-24: View of Reactor Building Looking Down . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-244 re 3.7.2-25: Enlarged View of Reactor Pool Looking Down . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-245 re 3.7.2-26: Extruded View of the NuScale Power Modules and Support Walls . . . . . . . . . . . . . 3.7-246 re 3.7.2-27: NuScale Power Module Model with Lug Restraint and Base Skirt Supports. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-247 re 3.7.2-28: NuScale Power Module Beam Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-248 re 3.7.2-29: Beam and Spring Model of Reactor Building Crane . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-249 re 3.7.2-30: Longitudinal Section View of Pool Water and NuScale Power Modules . . . . . . . . 3.7-250 re 3.7.2-31: Model of Reactor Building Pool Water. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-251 re 3.7.2-32: Half Sectional View of Reactor Building ANSYS Model with Pool Fluid and Backfill Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-252 re 3.7.2-33: ANSYS Model of Fluid, NuScale Power Modules, Foundation and Interior Pool Walls. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-253 re 3.7.2-34: 3D View of Pool without Water with 12 NuScale Power Modules . . . . . . . . . . . . . . 3.7-254 re 3.7.2-35: Plan View of Wall Segments used for FSI analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-255 re 3.7.2-36: Maximum Accelerations for X Wall Sections from ANSYS . . . . . . . . . . . . . . . . . . . . . 3.7-256 2 xviii Revision 1

re 3.7.2-37: Maximum Accelerations for Y Wall Sections from ANSYS . . . . . . . . . . . . . . . . . . . . . 3.7-257 re 3.7.2-38: Hydrodynamic Pressure for X Wall Sections from ANSYS . . . . . . . . . . . . . . . . . . . . . . 3.7-258 re 3.7.2-39: Hydrodynamic Pressure for Y Wall Sections from ANSYS . . . . . . . . . . . . . . . . . . . . . . 3.7-259 re 3.7.2-40: Maximum Accelerations for X Wall Sections, Soil Type 7 from SASSI2010 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-260 re 3.7.2-41: Maximum Accelerations for Y Wall Sections, Soil Type 7 from SASSI2010 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-261 re 3.7.2-42: Maximum Accelerations for X Wall Sections, Soil Type 8 from SASSI2010 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-262 re 3.7.2-43: Maximum Accelerations for Y Wall Sections, Soil Type 8 from SASSI2010 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-263 re 3.7.2-44: Maximum Accelerations for X Wall Sections, Soil Type 11 from SASSI2010 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-264 re 3.7.2-45: Maximum Accelerations for Y Wall Sections, Soil Type 11 from SASSI2010 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-265 re 3.7.2-46: Control Building (Looking Northeast) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-266 re 3.7.2-47: East-West Section Cut View of Control Building in Soil . . . . . . . . . . . . . . . . . . . . . . . . 3.7-267 re 3.7.2-48: North-South Section Cut View of Control Building in Soil . . . . . . . . . . . . . . . . . . . . . 3.7-268 re 3.7.2-49: Quarter View of Control Building in Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-269 re 3.7.2-50: SAP2000 Control Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-270 re 3.7.2-51: SAP2000 Control Building Model Beam Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-271 re 3.7.2-52: SAP2000 Control Building Model with Backfill Soil. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-272 re 3.7.2-53: Control Building SASSI2010 Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-273 re 3.7.2-54: Excavated Soil of Control Building SASSI2010 Model . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-274 re 3.7.2-55: Control Building SASSI2010 Model Backfill Soil Solid Elements . . . . . . . . . . . . . . . . 3.7-275 re 3.7.2-56: Control Building SASSI2010 Solid Elements Modeling the Basemat . . . . . . . . . . . 3.7-276 re 3.7.2-57: Control Building SASSI2010 Model Shell and Beam Element . . . . . . . . . . . . . . . . . . 3.7-277 re 3.7.2-58: Control Building SASSI2010 Model Beam Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-278 re 3.7.2-59: Structures and Backfill Soil of Triple Building SAP2000 Model . . . . . . . . . . . . . . . . . 3.7-279 re 3.7.2-60: Isometric View of South Side of Triple SAP2000 Model (Looking Northeast) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-280 re 3.7.2-61: Isometric View of South Side of Triple Building SAP2000 Model (Looking Northwest) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-281 re 3.7.2-62: Isometric View of North Side of Triple Building SAP2000 Model (Looking Southwest) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-282 2 xix Revision 1

re 3.7.2-63: Isometric View of Backfill Soil Elements around the Three Buildings . . . . . . . . . . . 3.7-283 re 3.7.2-64: Beam Elements of Triple Building SAP2000 Model. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-284 re 3.7.2-65: Spring or Link Elements of Triple Building SAP2000 Model. . . . . . . . . . . . . . . . . . . . 3.7-285 re 3.7.2-66: Elevation View of Triple Building SAP2000 Model Showing Separation . . . . . . . . 3.7-286 re 3.7.2-67: Isometric View of SASSI2010 Triple Building. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-287 re 3.7.2-68: North Half View of SASSI2010 Triple Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-288 re 3.7.2-69: SASSI2010 Triple Building Model Shown without Backfill . . . . . . . . . . . . . . . . . . . . . 3.7-289 re 3.7.2-70: SASSI2010 Triple Building Model Showing South Side of Three Buildings (Looking Northwest). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-290 re 3.7.2-71: SASSI2010 Triple Building Model Showing North Side of Three Buildings (Looking Southwest) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-291 re 3.7.2-72: Beam Elements of SASSI2010 Triple Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-292 re 3.7.2-73: Excavated Soil Solid Elements of the SASSI2010 Triple Building Model . . . . . . . . 3.7-293 re 3.7.2-74: Backfill Soil Solid Elements of the SASSI2010 Triple Building Model . . . . . . . . . . . 3.7-294 re 3.7.2-75: Rigid Soil Springs between Backfill and Free Field Soils of the SASSI2010 Triple Building Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-295 re 3.7.2-76: Interaction Nodes for Soil Impedance Calculation . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-296 re 3.7.2-77: Reactor Building Cracked Model - Modal Shape - (Mode 5). . . . . . . . . . . . . . . . . . . . 3.7-297 re 3.7.2-78: Reactor Building Cracked Model - Modal Shape - (Mode 6). . . . . . . . . . . . . . . . . . . . 3.7-298 re 3.7.2-79: Reactor Building Cracked Model - Modal Shape - (Mode 14) . . . . . . . . . . . . . . . . . . 3.7-299 re 3.7.2-80: Reactor Building Uncracked Model - Modal Shape - (Mode 5) . . . . . . . . . . . . . . . . . 3.7-300 re 3.7.2-81: Reactor Building Uncracked Model - Modal Shape - (Mode 8) . . . . . . . . . . . . . . . . . 3.7-301 re 3.7.2-82: Reactor Building Uncracked Model - Modal Shape - (Mode 15) . . . . . . . . . . . . . . . . 3.7-302 re 3.7.2-83: Control Building Cracked Model - Modal Shape - (Mode 40) . . . . . . . . . . . . . . . . . . 3.7-303 re 3.7.2-84: Control Building Cracked Model - Modal Shape - (Mode 49) . . . . . . . . . . . . . . . . . . 3.7-304 re 3.7.2-85: Control Building Cracked Model - Modal Shape - (Mode 81) . . . . . . . . . . . . . . . . . . 3.7-305 re 3.7.2-86: Control Building Uncracked Model - Modal Shape - (Mode 41) . . . . . . . . . . . . . . . . 3.7-306 re 3.7.2-87: Control Building Uncracked Model - Modal Shape - (Mode 49) . . . . . . . . . . . . . . . . 3.7-307 re 3.7.2-88: Control Building Uncracked Model - Modal Shape - (Mode 81) . . . . . . . . . . . . . . . . 3.7-308 re 3.7.2-89: Flow of Files among SASSI2010 Modules . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-309 re 3.7.2-90: Four I, J, K, and L Nodes and 1 through 5 Output Locations of Thick Shell Element . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-310 re 3.7.2-91: SASSI2010 Shear and Moment Resultants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-311 2 xx Revision 1

re 3.7.2-92: SASSI2010 Beam Element Local Axes for Forces and Moments . . . . . . . . . . . . . . . . 3.7-312 re 3.7.2-93: SASSI2010 Global Stresses at Centroid of a Solid Element. . . . . . . . . . . . . . . . . . . . . 3.7-313 re 3.7.2-94: Reactor Building Locations Selected for Relative Displacement . . . . . . . . . . . . . . . 3.7-314 re 3.7.2-95: Control Building Locations Selected for Relative Displacement . . . . . . . . . . . . . . . 3.7-315 re 3.7.2-96: Nodes Used for ISRS for NPM Bay Walls at the Pool Floor (EL. 25 0) . . . . . . . . . . . 3.7-316 re 3.7.2-97: NPM Lug Restraint Node Locations (EL. 71' 7"). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-317 re 3.7.2-98: Location of NPMs for 7 Module Case Study. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-318 re 3.7.2-99: Example ISRS from CSDRS compatible Time Histories for Soil Type 7 . . . . . . . . . . 3.7-319 re 3.7.2-100: Example ISRS from CSDRS compatible Time Histories for Soil Type 8 . . . . . . . . . . 3.7-320 re 3.7.2-101: Example ISRS from CSDRS compatible Time Histories for Soil Type 11 . . . . . . . . . 3.7-321 re 3.7.2-102: Example Combined and Enveloped ISRS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-322 re 3.7.2-103: Exampled Broadened ISRS at Multiple Damping Ratios . . . . . . . . . . . . . . . . . . . . . . . 3.7-323 re 3.7.2-104: Comparison between Standalone Reactor Building Model and Triple Building Model, ISRS at Northwest Corner, Top of Basement . . . . . . . . . . . . . . . . . . 3.7-324 re 3.7.2-105: Comparison between Standalone Reactor Building Model and Triple Building Model, ISRS at Northwest Corner, Top of Exterior Wall . . . . . . . . . . . . . . . 3.7-325 re 3.7.2-106: Comparison between Standalone Reactor Building Model and Triple Building Model, ISRS at Corner or Roof Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-326 re 3.7.2-107: Reactor Building ISRS for Floor at El. 24 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-327 re 3.7.2-108: Reactor Building ISRS for Floor at El. 25 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-328 re 3.7.2-109: Reactor Building ISRS for Floor at El. 50 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-329 re 3.7.2-110: Reactor Building ISRS for Floor at El. 75 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-330 re 3.7.2-111: Reactor Building ISRS for Floor at El. 100 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-331 re 3.7.2-112: Reactor Building ISRS for Floor at El. 126 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-332 re 3.7.2-113: Reactor Building ISRS for Roof at El. 181 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-333 re 3.7.2-114: ISRS at Reactor Building Crane Wheels at El. 145 6. . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-334 re 3.7.2-115: ISRS at NPM Bay Wall at the Pool Floor . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-335 re 3.7.2-116: ISRS at NPM Lug Restraints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-336 re 3.7.2-117: Control Building ISRS at Floor at El. 50 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-337 re 3.7.2-118: Control Building ISRS at Floor at El. 63 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-338 re 3.7.2-119: Control Building ISRS at Floor at El. 76 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-339 re 3.7.2-120: Control Building ISRS at Floor at El. 100 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-340 re 3.7.2-121: Control Building ISRS at Floor at El. 120 0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-341 2 xxi Revision 1

re 3.7.2-122: Control Building ISRS at Roof at El. 140 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-342 re 3.7.2-123: Comparison of 12 NPM and 7 NPM Model Results at Northwest Corner on Top of Basement (EL. 24-0). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-343 re 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) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-344 re 3.7.2-125: Comparison of 12 NPM and 7 NPM Model Results at Northeast Corner on Top of Basement (EL. 24-0). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-345 re 3.7.2-126: Comparison of 12 NPM and 7 NPM Model Results at Northwest Corner on Top of Roof Slab (EL. 181-0) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-346 re 3.7.2-127: Comparison of 12 NPM and 7 NPM Model Results at Mid-Span of Roof Slab (EL. 181-0) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-347 re 3.7.2-128: Comparison of 12 NPM and 7 NPM Model Results at Northwest Corner of Roof Slab (EL. 181-0) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-348 re 3.7.3-1: Bioshield. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-371 re 3.7.3-2: Conceptual Bioshield Vertical Face Plate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-372 re 3.7.3-3: Location In-structure Response Spectra Nodes for Design of Bioshields . . . . . . . 3.7-373 re 3.7.3-4: In-structure Response Spectra Used for the Evaluation of the Bioshield . . . . . . . 3.7-374 re 3.8.2-1: Containment Vessel Components and Building Elevations. . . . . . . . . . . . . . . . . . . . . 3.8-29 re 3.8.2-2: Passive Support Skirt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-30 re 3.8.2-3: Containment Vessel Lateral Lug Located within the NuScale Power Module Lug Restraints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-31 re 3.8.2-4: Containment Vessel Top Head Mechanical Penetrations . . . . . . . . . . . . . . . . . . . . . . . 3.8-32 re 3.8.2-5: Containment Vessel Top Head Instrumentation and Controls, Electrical, and Access Penetrations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-33 re 3.8.2-6: Typical Access Cover and O-Ring Seals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-34 re 3.8.2-7: Typical Non Secondary Side Containment Vessel Penetration Configuration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-35 re 3.8.2-8: Containment Vessel Reactor Pressure Vessel Support Boundary . . . . . . . . . . . . . . . 3.8-36 re 3.8.2-9: Containment Vessel Bottom Head Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-37 re 3.8.4-1: Reactor Building Concrete Structural Sections at First Floor (EL. 24'-0") . . . . . . . . . 3.8-84 re 3.8.4-2: Reactor Building Concrete Structural Sections at Second Floor (EL. 50'-0") . . . . . . 3.8-85 re 3.8.4-3: Reactor Building Concrete Structural Sections at Third Floor (EL. 75'-0") . . . . . . . . 3.8-86 re 3.8.4-4: Reactor Building Concrete Structural Sections at Fourth Floor (EL. 100'-0") . . . . . 3.8-87 re 3.8.4-5: Reactor Building Concrete Structural Sections at Fifth Floor (EL. 126'-0") . . . . . . . 3.8-88 re 3.8.4-6: Reactor Building Concrete Structural Sections at RBC (EL. 145'-6"). . . . . . . . . . . . . . 3.8-89 2 xxii Revision 1

re 3.8.4-7: Reactor Building Concrete Structural Sections at Roof (EL. 181'-0") . . . . . . . . . . . . . 3.8-90 re 3.8.4-8: Control Building Concrete Structural Sections at First Floor (EL. 50'-0") . . . . . . . . . 3.8-91 re 3.8.4-9: Control Building Concrete Structural Sections at Second Floor (EL. 76'-6") . . . . . . 3.8-92 re 3.8.4-10: Control Building Concrete Structural Sections at Third Floor (EL. 100'-0") . . . . . . . 3.8-93 re 3.8.4-11: Control Building Concrete Structural Sections at Fourth Floor (EL. 120'-0") . . . . . 3.8-94 re 3.8.4-12: Control Building Steel Framing of Roof to EL. 141' 2" . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-95 re 3.8.4-13: East-West (X) Longitudinal Hydrodynamic Load Regions . . . . . . . . . . . . . . . . . . . . . . 3.8-96 re 3.8.4-14: North-South (Y) Transverse Hydrodynamic Load Regions . . . . . . . . . . . . . . . . . . . . . . 3.8-97 re 3.8.4-15: Reactor Building SAP2000 Model (Looking Southwest) . . . . . . . . . . . . . . . . . . . . . . . . 3.8-98 re 3.8.4-16: Elevation View of Reactor Building SAP2000 Model Looking South. . . . . . . . . . . . . 3.8-99 re 3.8.4-17: Elevation View of Reactor Building SAP2000 Model Looking East . . . . . . . . . . . . . 3.8-100 re 3.8.4-18: Longitudinal Section View of Reactor Building SAP2000 Model . . . . . . . . . . . . . . . 3.8-101 re 3.8.4-19: Transverse Section View of Reactor Building SAP2000 Model . . . . . . . . . . . . . . . . . 3.8-102 re 3.8.4-20: Reactor Building Exterior Walls with 7000 psi and 5000 psi Concrete . . . . . . . . . . 3.8-103 re 3.8.4-21: Control Building SAP2000 Model With Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-104 re 3.8.4-22: Control Building SAP2000 Model Without Soil. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-105 re 3.8.4-23: Control Building SAP2000 Model View Looking West . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-106 re 3.8.4-24: Control Building SAP2000 Model View Looking East . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-107 re 3.8.4-25: Control Building SAP2000 Model View Looking North . . . . . . . . . . . . . . . . . . . . . . . . 3.8-108 re 3.8.4-26: Control Building SAP2000 Model View Looking South . . . . . . . . . . . . . . . . . . . . . . . . 3.8-109 re 3.8.4-27: Total Static Lateral Soil Pressure Distribution Reactor Building . . . . . . . . . . . . . . . . 3.8-110 re 3.8.5-1: SAP2000 Model for Evaluation of Design Forces in the Reactor Building Basemat Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-150 re 3.8.5-2: Static Base Pressure Contours for Governing Load Combination in the Reactor Building Basemat Model (Lb, in Units) . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-151 re 3.8.5-3: Seismic Base Pressure Contours from SASSI2010 Analysis in the Reactor Building Basemat Model (Lb, inch Units) . . . . . . . . . . . . . . . . . . . . . . . 3.8-152 re 3.8.5-3a: Dynamic Pressure Contours on Control Building Basemat (psi). . . . . . . . . . . . . . . . 3.8-153 re 3.8.5-4: M22 due to Static Base Pressure in the Reactor Building Basemat Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-154 re 3.8.5-5: M11 due to Static Base Pressure in the Reactor Building Basemat Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-155 re 3.8.5-6: M22 due to Seismic Base Pressure in the Reactor Building Basemat Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-156 2 xxiii Revision 1

re 3.8.5-7: M11 due to Seismic Base Pressure in the Reactor Building Basemat Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-157 re 3.8.5-8: RXB ANSYS Model with Backfill Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-158 re 3.8.5-9: Nonlinear Contact Region between Building and Soil . . . . . . . . . . . . . . . . . . . . . . . . 3.8-159 re 3.8.5-10: Nodes Selected for Settlement Values . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-160 re 3.8.5-11: RXB Skin Nodes on Backfill Soil Vertical Boundaries for Applying SASSI Acceleration Time Histories. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-161 re 3.8.5-12: RXB Foundation Bottom Skin Nodes for Applying SASSI Acceleration Time Histories. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-162 re 3.8.5-13: Displacements from SASSI Applied to ANSYS Model Boundary. . . . . . . . . . . . . . . . 3.8-163 re 3.8.5-14: Displacements from SASSI Applied to ANSYS Model Boundary. . . . . . . . . . . . . . . . 3.8-164 re 3.8.5-15: Nonlinear Contact Element between Backfill and Surrounding Soil . . . . . . . . . . . 3.8-165 re 3.8.5-16: Buoyancy Load on Basemat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-166 re 3.8.5-17: Soil Type 7 - Acceleration Time History - Vertical. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-167 re 3.8.5-18: Soil Type 7 - Acceleration Time History - E-W . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-168 re 3.8.5-19: Soil Type 7 - Acceleration Time History - N-S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-169 re 3.8.5-20: Soil Type 8 - Acceleration Time History - Vertical. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-170 re 3.8.5-21: Soil Type 8 - Acceleration Time History - E-W . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-171 re 3.8.5-22: Soil Type 8 - Acceleration Time History - N-S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-172 re 3.8.5-23: Soil Type 11 - Acceleration Time History - Vertical . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-173 re 3.8.5-24: Soil Type 11 - Acceleration Time History - E-W . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-174 re 3.8.5-25: Soil Type 11 - Acceleration Time History - N-S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-175 re 3.8.5-26: Nonlinear Contact Region between CRB and Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-176 re 3.8.5-27: CRB Time Histories from SASSI Applied to ANSYS Model Boundary. . . . . . . . . . . . 3.8-177 re 3.8.5-28: Soil Type 11, Capitola Input - Acceleration Time History - Vertical . . . . . . . . . . . . . 3.8-178 re 3.8.5-29: Soil Type 11, Capitola Input - Acceleration Time History - E-W. . . . . . . . . . . . . . . . . 3.8-179 re 3.8.5-30: Soil Type 11, Capitola Input - Acceleration Time History - N-S . . . . . . . . . . . . . . . . . 3.8-180 re 3.8.5-31: Soil Type 7, Capitola Input - Acceleration Time History - Vertical . . . . . . . . . . . . . . 3.8-181 re 3.8.5-32: Soil Type 7, Capitola Input - Acceleration Time History - E-W . . . . . . . . . . . . . . . . . . 3.8-182 re 3.8.5-33: Soil Type 7, Capitola Input - Acceleration Time History - N-S . . . . . . . . . . . . . . . . . . 3.8-183 re 3.8.5-34: CRB Skin Nodes on Backfill Outer Boundaries for Applying SASSI Time Histories . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-184 re 3.8.5-35: CRB Foundation Bottom Skin Nodes for Applying SASSI Time Histories. . . . . . . . 3.8-185 2 xxiv Revision 1

re 3.8.5-36: Buoyancy Load on Basemat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-186 re 3.8.5-37: Static Soil Pressure on CRB Outer Walls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-187 re 3.8.5-38: CRB Static Soil Pressure from Poisson's Ratio Effect - Soil Type 11 . . . . . . . . . . . . . 3.8-188 re 3.8.5-39: CRB Static Soil Pressure from Poisson's Ratio Effect - Soil Type 7 . . . . . . . . . . . . . . 3.8-189 re 3.8.5-40: CRB SAP2000 Model with Backfill Soil . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-190 re 3.8.5-41: SAP2000 Model for Settlement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-191 re 3.8.5-42: Total Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-192 re 3.8.5-43: Total Uncracked Base Vertical Reaction Time History due to Capitola for Soil Type 7. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-193 re 3.8.5-44: Total Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-194 re 3.8.5-45: Total Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 11. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-195 re 3.8.5-46: Total Cracked Base Vertical Reaction Time History due to Lucerne for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-196 re 3.8.5-47: Total Cracked Base Vertical Reaction Time History due to Lucerne for Soil Type 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-197 re 3.8.5-48: CRB Foundation Time History Location Designations . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-198 re 3.8.5-49: Reaction Force at Location A (S11 - Vertical Excitation) . . . . . . . . . . . . . . . . . . . . . . . 3.8-199 re 3.8.5-50: Relative Displacement (Uplift) at Location A (S11 - Vertical Excitation). . . . . . . . . 3.8-200 re 3.8.5-51: Lateral Relative Displacements (Sliding) at Location A (S11 - Vertical Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-201 re 3.8.5-52: RXB Foundation Time History Location Designations . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-202 re 3.8.5-53: Lateral Relative Displacements (Sliding) at Location A (S7 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-203 re 3.8.5-54: Lateral Relative Displacements (Sliding) at Location B (S7 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-204 re 3.8.5-55: Lateral Relative Displacements (Sliding) at Location C (S7 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-205 re 3.8.5-56: Lateral Relative Displacements (Sliding) at Location D (S7 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-206 re 3.8.5-57: Lateral Relative Displacements (Sliding) at Location A (S7 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-207 re 3.8.5-58: Lateral Relative Displacements (Sliding) at Location B (S7 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-208 2 xxv Revision 1

re 3.8.5-59: Lateral Relative Displacements (Sliding) at Location C (S7 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-209 re 3.8.5-60: Lateral Relative Displacements (Sliding) at Location D (S7 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-210 re 3.8.5-61: Lateral Relative Displacements (Sliding) at Location A (S11 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-211 re 3.8.5-62: Lateral Relative Displacements (Sliding) at Location B (S11 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-212 re 3.8.5-63: Lateral Relative Displacements (Sliding) at Location C (S11 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-213 re 3.8.5-64: Lateral Relative Displacements (Sliding) at Location D (S11 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-214 re 3.8.5-65: Lateral Relative Displacements (Sliding) at Location A (S11 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-215 re 3.8.5-66: Lateral Relative Displacements (Sliding) at Location B (S11 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-216 re 3.8.5-67: Lateral Relative Displacements (Sliding) at Location C (S11 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-217 re 3.8.5-68: Lateral Relative Displacements (Sliding) at Location D (S11 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-218 re 3.8.5-69: Lateral Relative Displacements (Sliding) at Location A (S8 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-219 re 3.8.5-70: Lateral Relative Displacements (Sliding) at Location B (S8 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-220 re 3.8.5-71: Lateral Relative Displacements (Sliding) at Location C (S8 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-221 re 3.8.5-72: Lateral Relative Displacements (Sliding) at Location D (S8 - E-W Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-222 re 3.8.5-73: Lateral Relative Displacements (Sliding) at Location A (S8 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-223 re 3.8.5-74: Lateral Relative Displacements (Sliding) at Location B (S8 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-224 re 3.8.5-75: Lateral Relative Displacements (Sliding) at Location C (S8 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-225 re 3.8.5-76: Lateral Relative Displacements (Sliding) at Location D (S8 - N-S Excitation) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-226 re 3.8.5-77: Total CRB Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 7. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-227 2 xxvi Revision 1

re 3.8.5-78: Total CRB Uncracked Base Vertical Reaction Time History due to Capitola for Soil Type 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-228 re 3.8.5-79: Total CRB Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 8. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-229 re 3.8.5-80: Total CRB Cracked Base Vertical Reaction Time History due to Capitola for Soil Type 11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-230 re 3.8.5-81: Total CRB Cracked Base Vertical Reaction Time History due to Lucerne for Soil Type 7. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-231 re 3.8.5-82: Total CRB Cracked Base Vertical Reaction Time History due to Lucerne for Soil Type 9. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-232 re 3.9-1: Nuscale Power Module Showing Reactor Vessel Internals Component Assemblies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-99 re 3.9-2: Upper Riser Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-100 re 3.9-3: Lower Riser Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-101 re 3.9-4: Core Support Assembly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.9-102 re 3B-1: Whitney Rectangular Stress Block. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-104 re 3B-2: SAP2000 Membrane and Sheer Force Definition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-105 re 3B-3: SAP2000 Bending Moment Definition. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-106 re 3B-4: SASSI2010 Membrane, Shear Force, and Bending Moment Definitions . . . . . . . . 3B-107 re 3B-5: SAP2000 Frame Element Results Definition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-108 re 3B-6: SASSI2010 Frame Element Results Definition. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-109 re 3B-7: SAP2000 Elevation View and Shell Element Numbers at RXB Grid Line 1 (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-110 re 3B-8: RXB Reinforcement Elevation at Grid Line 1 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-111 re 3B-9: RXB Reinforcement Section View of Wall on Grid Line 1. . . . . . . . . . . . . . . . . . . . . . . 3B-112 re 3B-10: SAP2000 Elevation View and Shell Element Numbers at RXB Grid Line 3 (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-113 re 3B-11: RXB Reinforcement Elevation at Grid Line 3 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-114 re 3B-12: RXB Reinforcement Section View of Pool Weir Wall on Grid Line 3. . . . . . . . . . . . . 3B-115 re 3B-13: RXB Reinforcement Section View of Stiffener Wall on Grid Line 3. . . . . . . . . . . . . . 3B-116 re 3B-14: SAP2000 Elevation View and Shell Element Numbers at RXB Grid Line 4 (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-117 re 3B-15: RXB Reinforcement Elevation at Grid Line 4 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-118 re 3B-16: RXB Reinforcement Section View of 5 ft Thick Wall on Grid Line 4 . . . . . . . . . . . . . 3B-119 re 3B-17: RXB Reinforcement Section View of 4 ft Thick Wall on Grid Line 4 . . . . . . . . . . . . . 3B-120 2 xxvii Revision 1

re 3B-18: SAP2000 Elevation View and Shell Element Numbers at RXB Grid Line 6 (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-121 re 3B-19: RXB Reinforcement Elevation at Grid Line 6 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-122 re 3B-20: RXB Reinforcement Section View of Upper Stiffener Wall on Grid Line 6 . . . . . . . 3B-123 re 3B-21: RXB Reinforcement Section Views of Pool Wall on Grid Line 6. . . . . . . . . . . . . . . . . 3B-124 re 3B-22: SAP2000 Elevation View and Shell Element Numbers at RXB Grid Line E (Looking North) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-125 re 3B-23: RXB Reinforcement Elevation at Grid Line E Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-126 re 3B-24: RXB Reinforcement Section View of Wall on Grid Line E. . . . . . . . . . . . . . . . . . . . . . . 3B-127 re 3B-25: SAP2000 Plan View and Shell Element Numbers on Slab at RXB EL 100-0. . . . . 3B-128 re 3B-26: RXB Reinforcement Plan at EL 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-129 re 3B-27: RXB Reinforcement Section View of Slab at EL 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . 3B-130 re 3B-28: SAP2000 Plan View and Shell Element Numbers on RXB Roof Slab . . . . . . . . . . . . 3B-131 re 3B-29: RXB Reinforcement Plan for Roof Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-132 re 3B-30: RXB Reinforcement Section View of Roof Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-133 re 3B-31: SAP2000 View and Frame Element Numbers of Pilasters on RXB Grid Line A Wall. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-134 re 3B-32: RXB Reinforcement Detail for Pilaster Type 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-135 re 3B-33: RXB Reinforcement Detail for Pilaster Type 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-136 re 3B-34: RXB Reinforcement Detail for Pilaster Type 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-137 re 3B-35: RXB Reinforcement Detail for Pilaster Type 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-138 re 3B-36: RXB Reinforcement Detail for Pilaster Type 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-139 re 3B-37: SAP2000 View and Frame Element Numbers of Beams on RXB EL 75'-0" Slab . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-140 re 3B-38: RXB Reinforcement Detail for Type 1 T-Beams at EL 75'-0" . . . . . . . . . . . . . . . . . . . . 3B-141 re 3B-39: RXB Reinforcement Detail for Type 2 T-Beams at EL 75'-0" . . . . . . . . . . . . . . . . . . . . 3B-142 re 3B-40: SAP2000 View and Frame Element Numbers of Buttresses at Grid Line 1 on RXB EL. 126'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-143 re 3B-41: RXB Reinforcement Detail for Buttress Type 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-144 re 3B-42: SAP2000 Elevation View and Shell Element Numbers for West Wing Wall at Grid Line 4. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-145 re 3B-43: RXB Reinforcement Elevation at RXB Grid Line 4 Wall . . . . . . . . . . . . . . . . . . . . . . . . . 3B-146 re 3B-44: RXB Reinforcement Section View of 5 Foot Thick Wall on RXB Grid Line 4. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-147 2 xxviii Revision 1

re 3B-45: SAP2000 Elevation View and Shell Element Numbers at RXB Wall at Grid Line B (Looking North). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-148 re 3B-46: RXB Reinforcement Elevation at RXB Wall at Grid Line B . . . . . . . . . . . . . . . . . . . . . . 3B-149 re 3B-47: RXB Reinforcement Section View of RXB Wall at Grid Line B . . . . . . . . . . . . . . . . . . . 3B-150 re 3B-48: Elevation View of the NPM Base Support at RXB Pool Floor . . . . . . . . . . . . . . . . . . . 3B-151 re 3B-49: Plan View of NPM Base Support Passive Support Ring . . . . . . . . . . . . . . . . . . . . . . . . 3B-152 re 3B-50: Plan View of NPM Base Support Bearing Plate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-153 re 3B-51: NPM Lug Support Plan View and Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-154 re 3B-52: NPM Lug Location . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-155 re 3B-53: NPM Lug Support SAP2000 Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-156 re 3B-54: NPM Lug Support SAP2000 Model Close-Up . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-157 re 3B-55: NPM Lug Support Liner Plate Section . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-158 re 3B-56: NPM Lug Support Liner Plate and Shear Lugs (Shown in Red) . . . . . . . . . . . . . . . . . 3B-159 re 3B-57: NPM Lug Support Model showing internal Stiffener Plates. . . . . . . . . . . . . . . . . . . . 3B-160 re 3B-58: NPM Lug Support Loading (W-Lug-PY+) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-161 re 3B-59: NPM Lug Support Loading (W-Lug-PY-) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-162 re 3B-60: NPM Lug Support SAP2000 Model Restraints. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-163 re 3B-61: Stiffener Plate Section Cut Groups (Fins) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-164 re 3B-62: S11 Stress plotted on the Deflected Shape due to Load Combination W-Lug-PY+ (psi). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-165 re 3B-63: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-166 re 3B-64: Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-167 re 3B-65: SAP2000 Elevation View and Shell Element Numbers at CRB Grid Line 3 (Looking North) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-168 re 3B-66: CRB Reinforcement Elevation at Grid Line 3 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-169 re 3B-67: CRB Reinforcement Section View of Wall on Grid Line 3. . . . . . . . . . . . . . . . . . . . . . . 3B-170 re 3B-68: SAP2000 Elevation View and Shell Element Numbers at CRB Grid Line 4 (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-171 re 3B-69: CRB Reinforcement Elevation at Grid Line 4 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-172 re 3B-70: CRB Reinforcement Section View of Wall on Grid Line 4. . . . . . . . . . . . . . . . . . . . . . . 3B-173 re 3B-71: SAP2000 Elevation View and Shell Element Numbers at Grid Line A (Looking West) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-174 re 3B-72: CRB Reinforcement Elevation at Grid Line A Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-175 re 3B-73: CRB Reinforcement Section View of Wall on Grid Line A . . . . . . . . . . . . . . . . . . . . . . 3B-176 2 xxix Revision 1

re 3B-74: CRB Basemat View of Finite Element Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-177 re 3B-75: CRB Reinforcement Plan of Basemat Foundation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-178 re 3B-76: Cross Section of CRB Basemat Showing Reinforcing Steel. . . . . . . . . . . . . . . . . . . . . 3B-179 re 3B-77: SAP2000 Plan View and Shell Element Numbers on CRB Slab at EL. 100'-0" . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-180 re 3B-78: CRB Reinforcement Plan at EL. 100-0 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-181 re 3B-79: CRB Reinforcement Section Views of Slab at EL. 100-0 . . . . . . . . . . . . . . . . . . . . . . . 3B-182 re 3B-80: SAP2000 View and Frame Element Numbers of Pilasters on CRB Grid Line 1 Wall . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-183 re 3B-81: CRB Reinforcement Detail for Pilaster Type 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-184 re 3B-82: CRB Reinforcement Detail for Pilaster Type 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-185 re 3B-83: SAP2000 View and Frame Element Numbers of T-Beams on CRB EL. 120'-0" Slab. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3B-186 re 3B-84: CRB Reinforcement Detail for T-Beam (Type 1) at EL. 120'-0" . . . . . . . . . . . . . . . . . . 3B-187 re 3B-85: CRB Reinforcement Detail for T-Beam (Type 2) at EL. 120'-0" . . . . . . . . . . . . . . . . . . 3B-188 re 3C-1: Containment Liquid Space Metal and Liquid Temperatures with Bounding Curve (Zones A and B) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-27 re 3C-2: Containment Vapor Space Metal and Gas Temperatures with Bounding Curve (Zones C, D, E, and F). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-28 re 3C-3: Bounding Envelope for Average Vapor Temperature at Top of Module (Zone G) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-29 re 3C-4: Bounding Envelope for Maximum Vapor Temperatures at Reactor Building El 145'-0 (Zone H) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3C-30 2 xxx Revision 1

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 1

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 1

including appropriate combinations of the effects of normal operating and accident conditions. The NuScale Power Plant's site design 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 1

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 Section 9.5 Other Auxiliary Systems 2 3.1-4 Revision 1

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 1

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 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 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.

2 3.1-7 Revision 1

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 ASME B&PV Code,Section III Division 1, Subsection NB (Reference 3.1-2) and the materials selected for fabrication of the RCPB meet the ASME B&PV Code,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 1

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 1

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 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 2 3.1-14 Revision 1

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 Power Plant design includes a remote shutdown station which has the necessary instrumentation and controls to maintain the NPM in a safe condition during hot shutdown and to bring the NPM to safe shutdown.

Conformance or Exception The NuScale Power Plant design conforms to GDC 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 Section 9.4.1 Control Room Area Ventilation System 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 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 2 3.1-15 Revision 1

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) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum 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 2 3.1-16 Revision 1

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 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. 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.

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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.

Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.

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.

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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 Section 4.6 Functional Design of Control Rod Drive System 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.

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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 Section 9.3 Process Auxiliaries 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, without margin for stuck rods, provided the specified acceptable fuel design limits for critical heat flux would not be exceeded by the return to power.

The CVCS, with boron addition, and CRDS are designed for a combined capability of controlling reactivity changes that assures the capability to cool the core under postulated accident conditions with margin for stuck rods as explained in Section 4.3.1.5. Conservative analysis indicates that a return to power could occur following a reactor trip under the condition that the highest worth CRA does not insert, coincident with the CVCS being unavailable. Consequently, the GDC is modified for the NuScale design to address the shutdown capability for postulated accidents.

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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.

Relevant FSAR Chapters and Sections Section 4.3 Nuclear Design Section 4.6 Functional Design of Control Rod Drive System 2 3.1-21 Revision 1

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 a ASME B&PV Code,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 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.

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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 B&PV Code,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 accident conditions (1) the boundary behaves 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 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.

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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 B&PV Code,Section III Division 1, Subsection NB Class 1 pressure-retaining components in accordance with ASME B&PV Code,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 changes in fracture toughness in accordance with "Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels,"

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.

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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 Section 9.3.4 Chemical and Volume Control System 2 3.1-25 Revision 1

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 Section 8.3 Onsite Power Systems 2 3.1-26 Revision 1

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.

The ECCS is designed such that no single failure prevents the system from performing its safety function including loss of onsite or offsite electrical power, initiation logic, and 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 2 3.1-27 Revision 1

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 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 B&PV Code,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 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 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.

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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 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 the system safety function can be accomplished, assuming a single failure.

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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 B&PV Code,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 inspection of the surfaces for fouling or degradation that could potentially impede heat transfer to the UHS.

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

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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, 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.

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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.

Conformance or Exception 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 2 3.1-32 Revision 1

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.

Implementation in the NuScale Power Plant Design 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.

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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.

The water level of the UHS is monitored by level instrumentation which provides a signal to the spent fuel pool cooling system for the addition of demineralized water as normal makeup when a low pool water level is detected.

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

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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.

Implementation in the NuScale Power Plant Design 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.

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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 Section 6.2 Containment Systems 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 2 3.1-36 Revision 1

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 B&PV Code 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.

Conformance or Exception 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 2 3.1-37 Revision 1

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 B&PV Code,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 MPS detects low AC voltage. The closure times are designed to minimize release of containment atmosphere to the environment.

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 2 3.1-38 Revision 1

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 consideration of the population density, use characteristics, and physical characteristics of the site environs.

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 2 3.1-39 Revision 1

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 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.

Isolation valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation valves shall be 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 2 3.1-40 Revision 1

with ASME B&PV Code 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 B&PV Code,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 B&PV Code,Section III Division 1, Subsection NB, Class 1 components.

The DHRS lines penetrate containment and are neither part of the RCPB nor connected directly to the containment atmosphere. The DHRS is a closed system inside and 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 B&PV Code,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.

2 3.1-41 Revision 1

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 Section 9.3 Process Auxiliaries Chapter 11 Radioactive Waste Management 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 2 3.1-42 Revision 1

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.

Chapter 11 describes the radioactive waste systems and the means provided to confine and filter radioactive material.

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 2 3.1-43 Revision 1

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.

Implementation in the NuScale Power Plant Design 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 2 3.1-44 Revision 1

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.

The primary coolant fluids are not required to be recirculated outside of containment following an accident. Radioactivity levels contained in the facility effluent and 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.

2 3.1-45 Revision 1

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, 2013 edition,Section III Divison 1, Subsection NB, Class 1 Components, New York, NY.

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

3.1-4 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2013 edition,Section III Division 1, Subsection NC, Class 2 Components, New York, NY.

3.1-5 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2013 Edition,Section XI Division 1, "Rules for Inservice Inspection of Nuclear Components," 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 3.1-46 Revision 1

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 non-safety 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 non-safety-related SSC where specific regulatory guidance applies (e.g.,

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

2 3.2-1 Revision 1

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.

2 3.2-2 Revision 1

1.1 Seismic Category I 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.

2 3.2-3 Revision 1

interaction with a Seismic Category I SSC exists. Additionally, non-safety related instrument lines from safety related pressure boundaries are required to maintain pressure integrity.

Seismic Category II SSC are subject to the pertinent quality assurance program requirements of 10 CFR 50, Appendix B as noted in Table 3.2-1.

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.

2 3.2-4 Revision 1

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.

The following subsections also describe the codes and standards applicable to supports for Quality Group A, B, C, and D components. The reactor vessel internals (see Section 3.9.5) 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 B&PV Code (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 B&PV Code (Reference 3.2-1). Supports for Quality Group A SSC meet the requirements for Class 1 supports in Section III, Division 1, Subsection NF of the ASME B&PV Code 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.

2 3.2-5 Revision 1

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 B&PV Code. Supports for Quality Group B SSC meet the requirements for Class 2 supports in Section III, Division 1, Subsection NF of the ASME B&PV Code and are not separately listed in Table 3.2-1.

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 B&PV Code. Supports for Quality Group C SSC meet the requirements for Class 3 supports in Section III, Division 1, Subsection NF of the ASME B&PV Code 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:

2 3.2-6 Revision 1

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

Pumps - Manufacturers standards

Atmospheric Storage Tanks - API-650 (Reference 3.2-5) or AWWA D-100 (Reference 3.2-6)

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, 2013 Edition,Section III, "Rules for Construction of Nuclear Facility Components," no addenda, New York, NY.

3.2-2 American National Standards Institute/Instrument Society of America (ANSI/

ISA)-67.02.01-1999, "Nuclear Safety-Related Instrument-Sensing Line Piping and Tubing Standard for Use in Nuclear Power Plants," November 1999.

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

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

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

2 3.2-7 Revision 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification 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 & 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/ANS 57.1-1992 N/A I

Top Support Structure

ASME NOG-1

Top Support Structure Diagonal Lifting Braces

NUREG-0554

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

FWS, Supply to SGs and DHR HXs FWIV

RCCWS, Inboard and Outboard Return and Supply

SGS, Steam Supply CIV/MSIVs and CIV/MSIV Bypasses 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

Feedwater plenums

Steam plenums

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

Upper and lower SG supports

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 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 Tier 2 3.2-8 Revision 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

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

(Note 4)

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, lower riser assembly, core support assembly, flow RXB A1 None Q None N/A I diverter, and pressurizer spray nozzles)

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 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 Check 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

PZR Spray Check Valve Hydrogen bottle and distribution assembly including excess flow valve RXB B2 None AQ-S None D II Tier 2 3.2-9 Revision 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

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

(Note 4)

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

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

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 Tier 2 3.2-10 Revision 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

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

(Note 4)

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

RG 1.52

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 II

RBV SFP Exhaust Filter Units, including fans RWB

RG 1.52 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

RG 1.52

Table 1 of SRP 11.5 LRWS, Liquid Radioactive Waste System Degasifiers RXB B2 None AQ None RW-IIa RW-IIa

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-IIb III

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 Tier 2 3.2-11 Revision 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

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

(Note 4)

SRWS, Solid Radioactive Waste System Spent Resin Storage Tanks RWB B2 None AQ None RW-IIa RW-IIa

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

RG 1.52

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 RPV Support Stand RXB A2 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 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)

Tier 2 3.2-12 Revision 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

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

(Note 4)

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 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.

Tier 2 3.2-13 Revision 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

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

(Note 4)

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

RG 1.52

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 Sampling 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

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 Tier 2 3.2-14 Revision 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

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

(Note 4)

MSS, Main Steam System

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

RXB Steam Traps

Secondary Main Steam Isolation Valves RXB B2 None AQ-S

Technical Specification Surveillance for D I

Secondary Main Steam Isolation Bypass Valves operability and in-service 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 RXB B2 None AQ-S Technical Specification Surveillance for D I operability and in-service testing.

Feedwater Supply Check Valve 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 in-service 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 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

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

(Note 4)

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

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

Effluent Radiation Transmitter

Discharge Flow Transmitter Tier 2 3.2-16 Revision 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

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

(Note 4)

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 and the piping CRB B2 None None None N/A II between the isolation devices 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 SAS, Service Air System All components CUB, ANB, B2 None None None N/A III RXB, TGB, RWB Tier 2 3.2-17 Revision 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification RG 1.29 or RG 1.143)

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

(Note 4)

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 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification 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, Non-Safety DC Electrical and AC Distribution 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 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification 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 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification 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 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification 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 I Traveling Jib Crane RXB B2 None N/A None N/A II Wet Hoist RXB B2 None AQ ASME NOG-1 N/A III RBCM, Reactor Building Components

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 D 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 1

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 (Ref.

(A1, A2, B1, B2) (A,B,C,D,E) Applicability (Note 3) Classification 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 non-safety 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.

Tier 2 3.2-23 Revision 1

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 1

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 continental United States as identified in RG 1.76, Rev. 1.

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

Translational speed . . . . . . . . . . . . . . . . . . . . . . . 46 mph 2 3.3-2 Revision 1

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 1

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, 2005.

2 3.3-4 Revision 1

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 1

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 1

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 1

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.

2 3.4-4 Revision 1

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.

2 3.4-5 Revision 1

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 a truck ramp on the south side of the Radwaste Building and the CRB tunnel.

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 identify the selected mitigation strategy for each room containing structures, systems, and components subject to flood protection.

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.

2 Protection of Structures Against Flood from External Sources 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 3.4-6 Revision 1

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 a truck ramp on the south side of the Radwaste Building and the CRB tunnel, which is below grade, 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% 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 design 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 groundwater and foreign material intrusion into the expansion gap between the end of the tunnel between the Reactor Building and the Control Building, and the 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 2 3.4-7 Revision 1

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))

The Seismic Category II portion of the CRB was analyzed along with the Seismic Category I portion of the structure and shown to be capable of withstanding the effects of the probable maximum precipitation.

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

2 3.4-8 Revision 1

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.

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

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

2 3.4-9 Revision 1

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 1

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 1

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 1

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.

2 3.5-1 Revision 1

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 1

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 1

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 1

core, is non-credible. The CRDM housing is a Class 1 appurtenance per ASME Section III. 1.3 Turbine Missiles The turbine generator building layout in relation to the overall site layout is shown on Figure 1.2-2. Safety related and risk significant SSC for the design are located principally within the RXB and CRB. The turbine generator rotor shafts are physically oriented such that the RXB and CRB are within the turbine low-trajectory hazard zone and considered to be unfavorably oriented with respect to the NPMs, as defined by RG 1.115, Revision 2. Safety-related and risk-significant SSC within the reactor and control building are protected from the effects of turbine missiles by limiting the generation of missiles from the turbine generators to be less than 10-5 consistent with Table 1 of RG 1.115. Item 3.5-1: A COL applicant that references the NuScale Power Plant certified design will provide a missile analysis for the turbine generator which demonstrates that the probability of a turbine generator producing a low trajectory turbine missile is less than 10-5. Section 10.2 describes the turbine generator requirements for turbine rotor integrity, including rotor material fracture toughness, overspeed protection, and inspection and testing. The turbine rotor inspection program along with the low probability of turbine missile generation provide assurance that safety related and risk significant SSC are protected from the adverse effects of turbine missiles, consistent with GDC 4. Item 3.5-2: A COL applicant that references the NuScale Power Plant certified design will address the effect of turbine missiles from nearby or co-located facilities. 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 3.5-5 Revision 1

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 design 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 3.5.1.4. 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 resistant to the other 2 3.5-6 Revision 1

Item 3.5-3: A COL applicant that references the NuScale Power Plant certified design 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. The RXB and CRB have not been credited to withstand turbine 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 FSAR 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 formulas discussed in "A Review of Procedures for 2 3.5-7 Revision 1

thicknesses calculated using the equations in this section for perforation and scabbing are increased by 20%. 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 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, 2 3.5-8 Revision 1

f'c = concrete compressive strength (lb/in2). 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. 2 3.5-9 Revision 1

results are addressed in Section 3.8.4. Design for impulsive and impactive loads is in accordance with ACI 349 "Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary," (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 3.5-10 Revision 1

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.

Steel 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. 2 3.5-11 Revision 1

3.5-1 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, "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). 3.5-2 American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2013 Edition with no Addenda (subject to the conditions specified in paragraph (b)(2) of section 50.55a). 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 (Amsterdam), V. 37, No. 2, 1976, pp. 183-203. 3.5-4 Cottrell, W.B., and Savolainen, A. W., "U.S. Reactor Containment Technology," ORNL NSIC-5, Ridge National Laboratory, Oak Ridge, TN: Volume 1, Chapter 6, 1965. 3.5-5 Russel, C. R., "Reactor Safeguards," New York: MacMillian, 1962. 3.5-6 American Concrete Institute, ACI 349, "Code Requirements for Nuclear Safety-Related Concrete Structures and Commentary," 2006. 3.5-7 American Institute of Steel Construction, AISC N690, "Specification for Safety-Related Steel Structures for Nuclear Facilities," 2012. 3.5-8 Bechtel Topical Report, BC-TOP-9A, Rev. 2, "Design of Structures for Missile Impact", by R.B. Linderman, J.V. Rotz, G.C.K. Yeh, September 1974. 3.5-9 Stephenson, A.E., "Full Scale Tornado Missile Impact Tests", EPRI NP440, Sandia Laboratories, Tonopa, NV, prepared for Electric Power Inst., Palo Alto, CA July 1977. 2 3.5-12 Revision 1

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 3.5-13 Revision 1

This section describes the design bases and measures needed to protect the reactor pressure vessel (RPV) and other essential systems and components inside or outside containment, including components of the reactor coolant pressure boundary, against the effects of pressurization, pipe rupture including jet impingement, pipe whip, and subcompartment pressurization resulting from a postulated rupture of piping located either inside or outside containment. 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 essential structures, systems, and components (SSC) are not impacted by the adverse effects of postulated piping rupture. 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. 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, and applicable Branch Technical Positions (BTPs). Section 3.6.1 identifies the high- and moderate-energy lines that have a potential to affect essential SSC, and describes the approaches used in the NuScale Power Plant design for protection of essential SSC. Section 3.6.2 provides the analytical methodology used to determine break locations and identifies the postulated breaks. 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. Finally, Section 3.6.5 describes the mitigation approaches used for postulated break locations if the dynamic consequences of the break cannot be tolerated. 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. This includes dynamic effects (pressurization, pipe whip and jet impingement) that may result from equipment failures and from events and conditions outside the nuclear power module. Plant designers are provided with options to address GDC 4 for high-energy line break (HELB) considerations. These options are as follows:

On a limited basis, portions of pipe may be excluded from HELB considerations provided they meet criteria regarding the design arrangement, stress and 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).

2 3.6-1 Revision 1

leak-before-break (LBB) analysis and is discussed in SRP 3.6.3. Leak-before-break 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.

High- and moderate-energy pipe systems that cannot be fully excluded using either the exclusion criteria of BTP 3-4, Section B.A.(ii) or LBB must be designed for HELB. The criteria for the specific locations for the postulated breaks are provided in BTP 3-4. 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 postulated rupture locations, the consequences of HELB can include pipe whip or jet impingement, both of which can potentially damage safety-related equipment required for safe shutdown. At break locations, the pipe must either be located such that there is no safety-related equipment in the area, or the pipe must be restrained from whip, and safety-related equipment protected from jet impingement, as needed. The piping systems that must be considered include the Class 1, 2, 3, and 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. 1.1 Identification of High- and Moderate-Energy Piping Systems 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.

By design, the NuScale Power Plant only has a small number of safety-related and risk-significant systems and components. These systems and components are primarily associated with the NPM, either inside the CNV, or mounted on the top of the CNV head. 2 3.6-2 Revision 1

the integrity of the containment isolation valves (CIVs) and decay heat removal Class 2 piping systems and condensers (outside of the CNV), including non-safety components credited in safety analyses
the emergency core cooling system (ECCS) valves
the decay heat removal system (DHRS) inside the CNV
the module protection system, including associated instruments, cables, and other components Figure 6.6-1 shows the lines that interface with the CNV.

The main steam and feedwater lines inside containment are part of the steam generator system in the NuScale system designation scheme. For the purpose of HELB analysis, these lines are referred to in relation to their process fluid system, as main steam system (MSS) and feedwater system (FWS) regardless of whether inside or outside of the CNV. The MSS and FWS are high-energy systems. The same practice is used for other systems that penetrate containment. The chemical and volume control system (CVCS) injection, discharge, pressurizer spray and high point vent lines are part of the RCS inside the CNV. The CVCS is a high-energy system. These lines are identified as the RCS injection, RCS discharge, pressurizer spray, and high point vent lines, or collectively as CVCS. The control rod drive system (CRDS) piping inside containment is functionally part of the moderate-energy reactor component cooling water system (RCCWS). The decay heat removal system (DHRS) piping is only associated with the NPM, and it is a high-energy system. The containment flooding and drain system (CFDS) is a single open pipe inside containment. This line is moderate-energy based on the amount of time in use. This line is identified as the containment system (CNTS) flooding and drain line both inside and outside the CNV. 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 SSC are consolidated into six groups. Each is discussed in a separate section.

inside the CNV (Section 3.6.1.1.1)
outside the CNV (to the disconnect flange) (Section 3.6.1.1.2)
in the Reactor Building (RXB), (outside the NPM disconnect flange)

(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) 2 3.6-3 Revision 1

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 comprehensive 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. 1.1.1 High- and Moderate-Energy Lines Inside the Containment Vessel There are ten high-energy lines inside the CNV: two main steam, two feedwater, RCS injection, RCS discharge, high point vent, pressurizer spray, and two DHRS condensate return. There are two moderate-energy lines inside the CNV, the CRDS cooling loop and CFDS (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 High- and Moderate-Energy Lines Outside the Containment Vessel (to the NuScale Power Module Disconnect Flange) The ten high-energy lines and two moderate-energy lines discussed in Section 3.6.1.1.1 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. Although not normally in use, this entire system is pressurized during NPM operation. 1.1.3 High- and Moderate-Energy Lines in the Reactor Building (outside the NuScale Power Module Disconnect Flange) Within the RXB, but outside the NPM disconnect flange, 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 the nominal diameter of the PSS lines, breaks do not need to be postulated in the PSS lines. The high-energy MSS and FWS lines exit the reactor pool through the North and South reactor pool walls, cross a mechanical equipment area and exit the RXB. Outside of the reactor pool bay, the high-energy CVCS lines run vertically downward in a pipe chase to the CVCS heat exchanger rooms at elevation 50' 0 and associated CVCS rooms at Elevations 24 0 and 35' 6". A break in any of these lines would only impact the function of the CVCS equipment for that module. The pipe chase can be seen on the general arrangement drawings in Section 1.2. The high-energy auxiliary boiler lines are routed to the module heatup heat exchangers in the CVCS rooms and to various service locations in the RXB. 2 3.6-4 Revision 1

1.1.4 High- and Moderate-Energy Lines 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 High- and Moderate-Energy Lines 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 High-Energy and Moderate-Energy Lines Outside the Reactor Building and Control Building Outside of the RXB and CRB there are four high-energy lines: MSS, FWS, auxiliary boiler, and extraction steam, and multiple moderate-energy lines (See Table 3.6-1). There is no 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 reactor pool bay, identify the location of high- and moderate-energy lines, and update Table 3.6-1 as necessary. 1.2 Types of Breaks 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 pipe whip, jet thrust reaction, jet impingement (dynamic effects), flooding, spray wetting, and increased temperature, pressure, and humidity (environmental effects). Through-wall leakage cracks are as defined in BTP 3-4, Revision 2, and are analyzed for localized flooding and environmental effects. For evaluation of spray wetting, flooding, and subcompartment pressurization effects, longitudinal breaks (with break flow areas equal to the piping flow area) are postulated in the main steam and feedwater piping outside the CNV. The dynamic effects of pipe whip and jet impingement are not evaluated for these breaks. 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 3.A. Moderate-energy lines are evaluated for through-wall leakage cracks and analyzed for flooding and environmental effects. The environmental effects of postulated moderate-energy leakage cracks are less severe than the inside containment and 2 3.6-5 Revision 1

1.3 Protection Methods Inside the CNV and reactor pool bay, including piping up to the pool wall, there is generally insufficient space to rely on distance (i.e., separation) or installation of traditional pipe whip restraints or jet shielding. Therefore, in these areas, the primary method employed by the NuScale design to mitigate the dynamic effects of pipe rupture is the integral shield restraint (ISR). This component is to be placed at all break locations identified in Figure 3.6-2 through Figure 3.6-15 and is discussed in more detail in Section 3.6.5. Currently, the ISR has been designed to be compatible with NPS 2 piping as all of the identified break locations are on NPS 2 piping. As the piping analysis is finalized other protection methods may be employed to protect against pipe whip and jet impingement, which may include equipment shields, barriers, and pipe whip restraints utilizing energy-absorbing structures. Pipe whip and jet protection methods other than the ISR are developed when postulated breaks are identified that cannot utilize an ISR. Outside of the reactor pool bay, protection is generally provided by separation in that there are a limited number of locations where safety-related or risk-significant equipment is co-located with high-energy lines. In those locations where they are in proximity to one another, an assessment is made to determine if the safety-related or risk-significant function is required to mitigate the consequences of the rupture. In general, the equipment is associated with the function provided by the line experiencing the break thus does not need to be protected, since the functionality is lost due to the break itself. If the dynamic effects of a break outside a reactor pool bay adversely affect safety-related or risk-significant systems or components, or could cause a transient in a second NPM, an ISR is installed or other conventional methods are used for shielding/restraint on the line with the postulated break. 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, and the methodology used to define the thrust at the postulated break location and the jet impingement loading on adjacent essential safety-related SSC. Pipe breaks on the MSS and FWS inside containment are replaced by small leakage cracks when the LBB criteria are applied (See Section 3.6.3). Jet impingement and pipe whip effects are not evaluated for these small leakage cracks. GDC 4 requires that SSC 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 by the adverse effects of postulated piping rupture, including pipe whip and jet impingement. Nonsafety-related systems are not required to be protected from the dynamic and environmental effects associated with the postulated rupture of piping. 2 3.6-6 Revision 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 nominal diameter; longitudinal breaks in fluid system piping that is 4-inch nominal diameter and greater, and leakage cracks in fluid system piping greater than 1-inch nominal diameter. Leakage cracks are also postulated in moderate-energy lines. The GDC 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. LBB is applied to the main steam and feedwater lines inside containment. This is discussed in Section 3.6.3. 2.1.1 Pipe Breaks Inside the Containment Vessel Due to the tight configuration and the concentration of safety-related and risk-significant SSC inside of the CNV, dynamic effects of pipe breaks cannot be tolerated. Therefore, the following strategies are employed for HELBs 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. See Figure 3.6-2, Figure 3.6-3, Figure 3.6-4 and Figure 3.6-5 for simplified drawings of the MSS and FWS high-energy piping inside containment. The CVCS RCS injection, RCS discharge, pressurizer spray, and high point vent lines inside containment are NPS 2, Schedule 160, ASME Class 1 stainless steel pipes. Due to their size, longitudinal breaks are not postulated. See Figure 3.6-6, Figure 3.6-7, Figure 3.6-8 and Figure 3.6-9 for simplified drawings of the CVCS high-energy RCS injection, RCS discharge, pressurizer spray, and high point vent lines, respectively, inside containment with postulated break locations indicated. 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 - The extremities of piping runs that connect to structures, components (e.g., vessels, pumps, valves), or pipe anchors, which act as rigid constraints to piping motion and thermal expansion 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 usage factor exceeds 0.4 2 3.6-7 Revision 1

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 - The extremities of piping runs that connect to structures, components (e.g., vessels, pumps, valves), or pipe anchors, which act as rigid constraints to piping motion and thermal expansion b) at intermediate locations selected by one of the following criteria: i) at each pipe fitting (e.g., elbow, tee, cross, flange, and nonstandard fitting), welded attachment, and valve. Or, where the piping contains no fittings, welded attachments, or valves, at one location at each extreme of the piping run adjacent to the protective structure. ii) at each location 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-3653. The DHRS condensate line inside containment runs from each feedwater line, just upstream of the feed plenum, to the containment upper cylindrical shell penetration. Breaks are postulated at the terminal ends as there are no intermediate welds in the piping run. See Figure 3.6-10 and Figure 3.6-11 for simplified drawings of the DHRS #1 and #2 high-energy lines inside and outside containment with postulated break locations indicated. The breaks are listed in Table 3.6-2. The CRDS 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 for the CNV. As a result, leakage cracks are not postulated further for the CRDS and CFDS lines inside containment. Final stress analysis will be performed concurrent with fabrication of the first NPM. The postulated break locations based upon the current analysis are listed in Table 3.6-2, and shown in Figure 3.6-6, Figure 3.6-7, Figure 3.6-8, Figure 3.6-9, Figure 3.6-10 and Figure 3.6-11. ITAAC A07, Pipe Break Hazards Protective Features Verification, was established to confirm the installation of protective features to mitigate 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 in the Reactor Pool Bay (Outside Containment) The containment isolation valves for the CVCS RCS injection, RCS discharge, pressurizer spray, high point vent line and the two feedwater lines are welded 2 3.6-8 Revision 1

Section B.A.(ii) have been applied to preclude the need for breaks to be postulated. In accordance with BTP 3-4 Section B.A.(ii), breaks are not postulated in this piping 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 reactor coolant system lines) from the CNV head to the first isolation valve 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. b) The cumulative usage factor is less than 0.1 unless environmentally assisted fatigue is considered in which case the usage factor is less than 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 Feedwater piping from containment to the first isolation valve does not exceed the following design 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 only one circumferential, and no longitudinal welds.

4) The length of the piping is the minimum practical.

5) There are no pipe anchors or restraints.

6) Guard pipes are not used.

2 3.6-9 Revision 1

Outboard of the containment isolation valves, the CVCS NPS 2, Schedule 160 RCS discharge, RCS injection, pressurizer spray, and high point vent 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. Breaks in these lines are postulated in accordance with BTP 3-4 Section B.A.(iii)(2) at the following locations: a) At terminal ends b) At intermediate locations selected by one of the following criteria: i) At each pipe fitting (e.g., elbow, tee, cross, flange, and nonstandard fitting), welded attachment, and valve. Or, where the piping contains no fittings, welded attachments, or valves, at one location at each extreme of the piping run adjacent to the protective structure. ii) At each location where stresses are calculated by the sum of equations (9) and (10) in NC/ND-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. Final stress analysis will be performed concurrent with fabrication of the first NPM. The postulated break locations based upon the current analysis are listed in Table 3.6-2, and shown in Figure 3.6-12 for the RCS injection line, Figure 3.6-12 for the RCS discharge line, Figure 3.6-13 for the pressurizer spray line, Figure 3.6-13 for the high point vent line, Figure 3.6-14 for the feedwater lines, and Figure 3.6-15 for the main steam lines. Due to the unique nature of the DHRS piping and the connections to the main steam line between the CIV and the CNV, these lines are specifically discussed in Section 3.6.2.5. Additionally, for the MS and FW piping, the break exclusion zone continues past the CIVs up to the penetrations at the reactor pool wall. These portions of piping are also discussed in Section 3.6.2.5. 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 in the reactor pool bay 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, Figure 3.6-12, Figure 3.6-13, Figure 3.6-14, and Figure 3.6-15 as appropriate. 2.1.3 Pipe Breaks in the Reactor Building (outside the Reactor Pool Bay) ASME Section III piping terminates at the NPM disconnect flanges or at the first valve outboard of the disconnect flange. Within the NPM, there are a large number 2 3.6-10 Revision 1

generally addressed with installation of an ISR device, as discussed in Section 3.6.1.3. Beyond the NPM disconnect flanges, 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 include safety-related and risk significant SSC, SSC that are credited in safety-related evaluations, and SSC that must remain operable to prevent the propagation of a more serious event. It is appropriate, therefore, for locations beyond the NPM disconnect flanges, to identify the target SSC that must be protected against the dynamic effects of postulated breaks, and then to identify vicinity high- and moderate-energy piping systems, determine the postulated rupture locations in those systems, and determine if protection is required. As fluid jets have the potential to impact SSC further away than pipe whip, a conservative approach is to evaluate ruptures of high- or moderate-energy piping located within 25 pipe diameters of the target SSC (Appendix A of SRP 3.6.2, Revision 3 Draft). Pipe routing for the balance-of-plant (BOP) (beyond the NPM disconnect flanges) is finalized with fabrication of the first NPM. The postulated BOP pipe routing is shown in Figure 3.6-17 for the large-bore feedwater lines, and Figure 3.6-16 for the large bore main steam lines. Similarly, the NuScale pipe rupture hazards analysis is completed for BOP high- and moderate-energy piping with finalization of the pipe routing outboard of the NPM disconnect flanges. The NuScale pipe rupture hazards analysis for BOP piping provides a summary of the analyses applicable to high- and moderate-energy pipe breaks, including:

identification of the high- and moderate-energy BOP piping systems, including line sizes, and location within the plant.
a list of target SSC that require protection based on systems identified as required to achieve safe shutdown. Additionally, SSC that provide for the continued safe operation of other NPMs are evaluated to ensure the postulated pipe rupture does not generate a more serious plant condition by initiating an operational occurrence or accident in another NPM.
at each location where a target SSC that requires protection is located, break locations and break types are postulated in the vicinity of high- and moderate-energy piping in accordance with Section 3.6.2.1.2.
sketches showing the locations of the postulated pipe ruptures, including identification of longitudinal and circumferential breaks and vicinity essential SSC.

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following criteria: at each pipe fitting (e.g., elbow, tee, cross, flange, and nonstandard fitting), welded attachment, and valve. Or, where the piping contains no fittings, welded attachments, or valves, at one location at each extreme of the piping run adjacent to the protective structure. at each location where stresses are calculated by the sum of equations. (9) and (10) in NC/ND-3653 of ASME Code, Section III, to exceed 0.8 times the sum of the stress limits given in NC/ND-3653, as delineated in BTP 3-4.

identification of the protective measures taken to mitigate the effects of postulated pipe failures for each identified essential SSC. Mitigation can include installation protecting features such as pipe whip restraints, equipment shields and ISRs. Evaluation of the unconstrained pipe whip, jet impingement, spray wetting, flooding, and other adverse environmental effects within the zone of influence may also be performed to show that the effects of the rupture are acceptable without mitigation.
for installed protective features, calculations of the imposed loads and stresses, evaluation of the unmitigated spray loads, zone of influence, installation characteristics, etc.
an evaluation and disposition of multi-module impacts in common pipe galleries. Pipe break on a high or moderate energy system associated with one NPM should not impair the ability of another NPM to continue normal operation, safe shutdown, maintenance, outage operations, or construction.
a conclusion that the affected NPM can be safely shut down and maintained in a safe shutdown following a pipe break.

A break in a high-energy MSS, FWS, or auxiliary boiler line in the RXB (outside of the Reactor Pool Bay) could potentially cause breaks or leakage cracks in lines of other modules, introducing an additional transient in a second module. Therefore, break locations are postulated and ISRs or other mitigating features may be used at the postulated break locations in the RXB (outside the Reactor Pool Bay) areas. Breaks are postulated in accordance with BTP Section B.A.(iii)(2) as shown above. The CVCS lines in the RXB (outside of the Reactor Pool Bay) are not co-located with any essential SSC. Therefore, dynamic effects are not a concern 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 and design appropriate protection features. This includes an evaluation and disposition of multi-module impacts in common pipe galleries, the identification of new detection and auto-isolation functions for mitigating an auxiliary boiler high-energy line break, and evaluations regarding subcompartment pressurization. The 2 3.6-12 Revision 1

2.1.4 Pipe Breaks in the Control Building There are no high-energy lines in the CRB. No breaks are postulated. Leakage cracks are postulated in the moderate-energy lines for flooding and environmental evaluations in Section 3.4 and 3.11, respectively. 2.1.5 Pipe Breaks in the Radioactive Waste Building There are no high-energy lines and essential equipment in the RWB. Therefore, no breaks or leakage cracks are postulated. 2.1.6 Pipe breaks On-site (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. 2.1.7 Moderate-Energy Line breaks and High-Energy Leakage Cracks Moderate-energy line breaks and high-energy leakage cracks do not cause dynamic effects. They are potential sources of environmental effects, (spray, flooding, pressurization, heatup, and radioactivity.) Within the CNV the limiting environmental conditions come from design basis accidents that result in ECCS actuation. The effect due to postulated specific moderate-energy line breaks are bounded by the effects of the main steam line breaks. These conditions are used for the evaluations in Section 3.11, Environmental Qualification of Mechanical and Electrical Equipment. Therefore specific moderate-energy breaks are not postulated here. Outside of the Reactor Pool Bay, environmental conditions are based upon the rupture of worst (typically largest or hottest) line in the proximity of safety-related SSC. For the environmental and flooding analysis, breaks are assumed to occur anywhere in both the high-energy and moderate-energy lines. 2.2 Guard Pipe Assembly Design Criteria Containment penetrations are fabricated as part of the CNV. Piping and components are either welded or bolted to the penetration nozzle. Guard pipes are not used within the CNV to the top of module area. 2 3.6-13 Revision 1

Within the CNV and reactor pool bay, all postulated break locations identified in Figure 3.6-2 through Figure 3.6-15 are addressed with an ISR. With the use of an ISR at each postulated break location, the pipe whip and jet impingement forces are effectively contained. The ISRs hold the pipe in place and the dynamic effects jet force and spray is mitigated by the ISR as described in Section 3.6.5.2. Accordingly, vicinity SSC are only minimally affected (minor localized flooding from drainage). Further evaluation of jet impingement forcing functions and response models for impact assessments is therefore not necessary. 2.4 Dynamic Analysis Methods to Verify Integrity and Operability As discussed in Section 3.6.2.3, all postulated break locations identified in Figure 3.6-2 through Figure 3.6-15 in the CNV and reactor pool bay are addressed with installation of an ISR. The ISR holds the pipe in place and the jet force and spray mitigated by the ISR as described in Section 3.6.5. Integrity and operability of vicinity SSC and building structures is therefore not challenged, and dynamic analysis to verify integrity and operability is not necessary. 2.5 Implementation of Criteria Dealing with Special Features Main Steam and Feedwater Lines from Containment to the Penetrations at the Reactor Pool Wall (including Tees to the Decay Heat Removal System) In accordance with BTP 3-4 Section B.A.(ii), breaks are not postulated in these segments of piping 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 main steam and feedwater lines do not exceed the following design stress and fatigue 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). The portions of the MS and FW lines beyond the CIVs also include flexible piping joints (i.e., ball joints) which significantly reduce thermal stress in the break exclusion zone. 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. Other welded features (thermowells, branch lines) within the piping segments have been minimized and are qualified using detailed stress analysis.

3) Piping welds are minimized to the extent possible. There are two eight-inch branch connections per MSS train for the DHRS lines. This results in three circumferential welds on each NPS 12 main steam line. There is a weld between the CNV safe-end 2 3.6-14 Revision 1

Figure 3.6-15). For both the MSS and FWS lines beyond the CIVs, piping welds have been minimized to the extent possible. Pipe bends are used instead of welded fittings where space allows.

4) The length of the piping is the minimum practical.

5) There are no pipe anchors or restraints welded to the surface of the pipe.

6) Guard pipes are not used.

7) The welds are included in the ISI program as described in Section 6.6.

Even though portions of the MS and FW lines are within the break exclusion zone, environmental effects resulting from the rupture of these lines is still considered as discussed in Section 3.6.1.2. 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 only 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 main steam line between the CNV and the main steam 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 permanently attached to the outside of the CNV at approximately the 50' level. The condenser is designed 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. Figure 3.6-10 and Figure 3.6-11 provide a simplified drawing of the DHRS. 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. In accordance with BTP 3-4 Section B.A.(ii), breaks are not postulated in this segment of piping because it meets the design criteria of the Section III of the ASME Boiler and Pressure Vessel Code, Subarticle NE-1120 and the following seven criteria: 2 3.6-15 Revision 1

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. Other welded features (thermowells, branch lines) within the piping segment have been minimized and are qualified using detailed stress analysis.

3) There are no longitudinal welds in this piping. Circumferential welds have been minimized to the extent possible. Piping bends are used in place of welded fittings where space allows.

4) The length of the piping is the minimum practical, considering that bends and jogs have been added to reduce the thermal stresses in the system.

5) There are no pipe anchors or restraints welded to the surface of the pipe.

6) Guard pipes are not used.

7) The welds are included in the ISI program as described in Section 6.6.

3 Leak-Before-Break Evaluation Procedures The GDC 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 main steam lines and feedwater lines inside containment are shown in Figure 3.6-2, Figure 3.6-3, Figure 3.6-4, and Figure 3.6-5, respectively. 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. 2 3.6-16 Revision 1

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 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 combination with water chemistry control, provide assurance that wall thinning by erosion-corrosion does not occur in the piping. 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 present, stress corrosion-cracking (SCC) is possible. The three conditions are:

There must be a corrosive environment.
The material itself must be susceptible.

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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. SA-312 TP304/304L dual certified stainless steel is also resistant to SCC given adequate control of dissolved oxygen levels. The alloy contains 0.03 maximum weight percent carbon, which mitigates sensitization. The use of cold worked austenitic stainless steels is generally avoided; however, if used, the yield strength as determined by the 0.2 percent offset method does not exceed 90 ksi. 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 2 3.6-18 Revision 1

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 B&PV Code 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. High-cycle Fatigue 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. (See Figure 3.6-2, Figure 3.6-3, Figure 3.6-4 and Figure 3.6-5) 2 3.6-19 Revision 1

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.
Also, being inside the CNV precludes fires, external missiles, or damage from moving heavy loads.
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. 2 3.6-20 Revision 1

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, tee-to- 4.563 0.500 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. The operating pressure and temperature for the FWS piping are 525 psia and 300 degrees F, respectively. 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. 2 3.6-21 Revision 1

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 (5) o = y/E (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: 2 3.6-22 Revision 1

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 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. 3.3.1.1 Algebraic Sum Method 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. 3.6-1 M i, Combined = S M ( M i, DW + M i, TH + M i, PR + M i, SSE + M i, SAM ) Eq. 3.6-2 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. 2 3.6-23 Revision 1

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. 3.6-3 M i, Combined = M i, DW + M i, TH + M i, PR + M i, SSE + M i, SAM Eq. 3.6-4 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. 3.6-5 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: F = F DW + F TH + F PR M X = ( M X ) DW + ( M X ) TH Eq. 3.6-6 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. 3.6-7 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. 3.6-8 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. 2 3.6-24 Revision 1

M = MX2 + MY2 + MZ2 Eq. 3.6-9 In the above equations, the moment due to the internal pressure is not included although it is included in Eq. 3.6-2 and Eq. 3.6-4, because the moment due to internal pressure is negligible. For limit load analysis, the thermal expansion and SAM loads are not included in Eq. 3.6-50 because they are secondary loads. The stresses due to axial loads and moments are then calculated by:

                                              = F         M
                                                    --- + -----                     Eq. 3.6-10 A Z where, A = cross-sectional area, Z = section modulus, M = moment, and F = axial force.

3.3.3 Leak Rate and Leakage Flaw Size Calculation 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 since 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. 3.6-11. n

                                           ----- = ------ +   ------

0 0 0 Eq. 3.6-11 2 3.6-25 Revision 1

          = true strain, 0 = reference strain (given by ---- ),

E E = Youngs modulus (psi),

          = true stress (psi),

0 = reference stress (the ASME Code-specified 0.2% 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. 3.6-12 E 1b t where, a = Rm = half crack length at the mean radius, Eq. 3.6-13 b = R m = half pipe circumference, Eq. 3.6-14

               = half crack angle in radians, R = mean pipe radius, R = R m ,

E = modulus of elasticity. B = MR

                                         ---------  = remote bending stress            Eq. 3.6-15 I

M = remote bending moment. 1 I = --- R - R R t 4 4 3

                                                              = area moment of inertia Eq. 3.6-16 4  0           i R0, Ri = pipe outer and inner radius, 2                                         3.6-26                                        Revision 1

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. 3.6-17 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. 1 M 0 = 4 0 R t cos

                                     --- - --- sin 2
                                                        = Reference bending moment                    Eq. 3.6-18 2 2 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. 3.6-19 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. 3.6-20 2Rt 2 3.6-27 Revision 1

R t where, 3 5 7 2 2 2 F T ( ) = 1 + 7.5 --- - 15 --- + 33 --- Eq. 3.6-22 3 5 7 2 2 2 F B ( ) = 1 + 6.8 --- - 13.6 --- + 20 --- Eq. 3.6-23 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. 3.6-24 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. For pure bending 1 B a R M n p = 0 aH --- ,n, --- ------- n Eq. 3.6-25 2 b t M 0 For pure tension 1 T a R F n p = 0 aH --- ,n, --- ------ n Eq. 3.6-26 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. 3.6-17. A different correction is needed for the combined tension and bending case because the plastic contributions from pure tension and pure bending 2 3.6-28 Revision 1

1n B a R M n p = 0.5 ( + ) 0 aH 2 --- ,n ,--- -------- Eq. 3.6-27 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. 3.6-24. 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. 3.6-28 The perimeter of an ellipse can be approximated by P wetted [ 3 ( a + /2 ) - ( 3a + /2 ) ( a + 3/2 ) ] Eq. 3.6-29 The hydraulic diameter is then calculated by 4A D H = ----------------- Eq. 3.6-30 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. 3.3.3.2 Two-phase Critical Flow Model 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 2 3.6-29 Revision 1

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. 3.6-31 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. 3.6-32 where, Gc    = mass flux of the fluid at the crack exit plane, c

So - Sl x e = ------------------- - = equilibrium fluid quality Eq. 3.6-33 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, c Sg = entropy of the saturated vapor at the crack exit plane pressure, 20x e , if x e <0.05 N1 = Eq. 3.6-34 1.0, if x e 0.05 L

                                                                                         - B  ------a- - 12 DH xc = N1 xe 1 - e                                           Eq. 3.6-35 La   = flow-path length, 4  Crack Opening Area D H = ------------------------------------------------------------
                                                                          -  = the hydraulic diameter perimeter (see Crack Opening Perimeter Eq. 3.6-30),

B=0.0523 = a constant based on experiments used in calculating exponential mixture quality relaxation, 2 3.6-30 Revision 1

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. 3.6-36 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. 3.6-37 Di 2 x = average fluid quality, v g = average specific volume of saturated vapor, v l = average specific volume of saturated liquid, DH -2 f = 2log ------- + 1.74 = Von Karman friction factor Eq. 3.6-38 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. 3.6-39 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 3.6-31 Revision 1

2 protrusions Eq. 3.6-40 2 G = average mass flux G2 of the fluid e v = e n L a = the total loss coefficient over the flow path Eq. 3.6-41 en = the number of velocity heads lost per unit flow path length, which is given in Eq. 3.6-43. Eq. 3.6-32 and Eq. 3.6-31 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 0 < ------- < 0.1 L G G - L

                                                     - ------- - 0.1 ,
                         =   L + ------------------                  0.1  -------  10 Eq. 3.6-42 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

2 3.6-32 Revision 1

en , 0 < ------- < 0.1 L G en L e n = e - ------ Eq. 3.6-43 n L 11- ------- - 0.1 , 0.1 ------- 10 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. 3.6-44 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.

3.3.3.3 Detectable Leak Rate The leakage of the piping systems inside the CNV can be detected by either using the CNV pressure sensor or the containment evacuation system (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 2 3.6-33 Revision 1

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. 3.6-45 2 f

                           -------- sin  + ( S m ) ( P m ), if  +  >

where, P m 0.5 ( - ) - ---------- , if + f

                            =                                                           Eq. 3.6-46 P m
                                         - ---------- , if  +  >

f F P m = -----x = primary membrane stress Eq. 3.6-47 A Fx = total applied axial force, A = cross-section area, a

    = ------c- = postulated through-wall circumferential crack half-angle Eq. 3.6-48 Rm Rm = pipe mean radius, SM = 1 = safety margin on the load, f = 0.5 (  y +  u )           = flow stress           Eq. 3.6-49 y = yield strength, and u = ultimate strength.

2 3.6-34 Revision 1

SI = SM ( Pm + Pb ) Eq. 3.6-50 where, M Rm P b = ---------------- = primary bending stress Eq. 3.6-51 I F

                                          - x I max -----         A M = -------------------------------- = applied maximum moment Eq. 3.6-52 Rm max = applied maximum stress, and I = area moment of inertia.

The max can be determined by making SI in Eq. 3.6-45 equal to that in Eq. 3.6-50. 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 moment is determined for a critical flaw size of twice the leakage flaw size. The margin of 2 on flaw size shall be satisfied.

5) calculate the low normal stress and corresponding maximum stress using the axial force and the bending moments by Eq. 3.6-10 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 2 3.6-35 Revision 1

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. 3.4.1.1 NPS 8 Straight Pipe Base Metal 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. 2 3.6-36 Revision 1

Eq. 3.6-8 and Eq. 3.6-9. 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. 3.6-10. 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. 3.4.1.3 NPS 8 Pipe-to-Safe-End Weld 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 since their weld material and dimensions are identical. The SBAC 2 3.6-37 Revision 1

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. 3.4.2 Analysis of Feedwater Piping 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 2 3.6-38 Revision 1

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. 3.6-6 and Eq. 3.6-7, the normal axial force and moment are calculated. The maximum axial force and moment are calculated using Eq. 3.6-8 and Eq. 3.6-9. 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. 3.6-10. 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. The SBAC is developed using the method described in Section 3.6.3.3.5 and plotted in Figure 3.6-30. The stress points are below the SBAC, demonstrating 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 2 3.6-39 Revision 1

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 data for SBAC are summarized in Table 3.6-3a. 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 data for SBAC are summarized in Table 3.6-3b. 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 3.6-29, Figure 3.6-30, Figure 3.6-31 and Figure 3.6-32. Therefore, it is concluded that the FWS piping meets the LBB criteria. 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 2 3.6-40 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, since 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. The LBB leak detection availability and limits will be included in the owner-controlled requirements manual. 4 High Energy Line Break Evaluation (Non-LBB) The GDC 4 requires that components be appropriately protected against the dynamic effects that may result from pipe ruptures. High-energy and moderate-energy piping systems that cannot be fully excluded using either the BTP 3-4, Section B.A.(ii) criteria, or LBB, must be designed for HELB. The specific locations for the postulated break locations are determined using the criteria in BTP 3-4. In general, welds 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 postulated rupture locations, the consequences of HELB can include pipe whip or jet impingement, both of which can potentially damage safety related equipment required for safe shutdown. At break locations, the pipe must either be located such that there is no 2 3.6-41 Revision 1

The piping systems that must be considered include the Class 1, Class 2, Class 3 and B31.1, high-energy and moderate-energy systems, located inside and outside of the CNV up to the reactor pool wall penetrations. Piping outside of the NPM is the responsibility of the COL applicant. 4.1 Postulation of Pipe Breaks in Areas Other than Containment Penetration With the exceptions of those portions of piping identified in the first two paragraphs of Section 3.6.2.1.2 and the portions of piping identified in Section 3.6.2.5, postulated pipe break locations are determined using the criteria described in Section 3.6.2. As a result of piping reanalysis, the highest stress locations may be shifted; however, the initially determined intermediate break locations need not be changed unless one of the following conditions exists: i) the dynamic effects from the new (as-built) intermediate break locations are not mitigated by installation of ISRs. ii) a change is necessary in pipe parameters such as major differences in pipe size, wall thickness, and routing. 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 seismically analyzed non-ASME Class piping are addressed in Section 3.6.2.1.3. 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 effect on the structure, irrespective of the fact that the criteria described in BTP 3-4, Part B, Items A(iii)(1) through (3) might not require the postulation of a break at that location. 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 2 3.6-42 Revision 1

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. The bases for break exclusion zones in areas away from containment penetrations and areas within containment penetrations are described in Section 3.6.2.1.2. The guidance relates to both Class 1 and Class 2 piping systems, where the allowable stresses identified in Section 3.6.2.1.2, which are itemized in Table 3.6-6 and Table 3.6-7, are based upon limits given in Table 3.6-5. The guidance is derived from BTP 3-4 Section B.A.(ii). 4.4 Jet Loads and Piping Moments Jet loads have been calculated for NPS 2 through NPS 12 for the CVCS, FWS, MSS, and DHRS process piping using guidance in SRP 3.6.2 and BTP 3-4. All piping runs generally employ 5D (i.e., five diameters) radius bends, with several larger radius bends (greater than 24 inch). Nonetheless, CVCS, FWS, MSS, and DHRS jet loads from 5D bends, assuming a pipe support near one end of the bend, result in creating a fully plastic hinge (i.e., plastic cross-section) as demonstrated in Table 3.6-8. Creation of a plastic hinge with jetting fluid has the potential for causing pipe whipping, as well as potential jet impingement on nearby essential equipment. HELB jet loads and associated maximum bending moments that occur on the supported-end of a 5D bend pipe when subjected to operating temperature and pressure conditions are shown in Table 3.6-8. In accordance with SRP 3.6.2, the jet thrust load is based on operating pressure and temperature (see Table 3.6-4). A 5D length moment-arm is utilized to determine if the lower-bound values result in creating plastic hinges. The second column from the right shows the value of the bending moment that would cause a fully-plastic cross-section. As evident from the right-most column of Table 3.6-8 for R = M / Mp, which is the ratio of maximum bending moment to fully-plastic bending moment, values are greater than unity. This implies that the high-energy lines postulated for HELB are subject to pipe whip. Further, piping runs with larger bend radii than 5D automatically are subject to pipe whip. Lastly, at locations with ISRs employed, a full- circumference pipe rupture at a weld might still occur, but the joint is restrained from moving further apart than the tolerance between the welded pipe collars and the ISR grooves (See Section 3.6.5). As such, typical ISRs allow a 0.125" gap at weld failure, such that a disk-type jet is developed (see Figure 3.6-35). Standard ANS 58.2 contains jet impingement force models for full-circumference break with limited separation. (Reference 3.6-15) 4.5 Break Locations inside the Containment Vessel As discussed in Section 3.6.2.1.1, only the CVCS and DHRS process piping are subject to HELB inside the CNV. The MSS and FWS are qualified for LBB (Section 3.6.3), thus the welds in MSS and FWS are excluded from break dynamic effects. 2 3.6-43 Revision 1

Table 3.6-2 identifies the resultant postulated break locations within the high-energy CVCS reactor coolant system discharge, RCS injection, pressurizer spray, RCS high-point degasification, and DHRS lines inside containment. To preclude the need to further evaluate the consequences of pipe whip and jet impingement at these locations, NuScale ISRs are installed (see Section 3.6.5). 4.6 High-Energy Piping outside Containment Table 3.6-2 identifies the postulated break locations within the high-energy CVCS RCS discharge, RCS injection, PZR spray, and RCS high-point degasification lines outside containment. To preclude the need to further evaluate the consequences of pipe whip and jet impingement at these locations, NuScale ISRs are installed (see Section 3.6.5). Piping outside of the NPM is the responsibility of the COL applicant. 5 Integral Jet Impingement Shield and Pipe Whip Restraint One method used in the NuScale design to mitigate the dynamic effects of a pipe rupture is installation of an ISR. The basic design of the ISR is a cylindrical sleeve which encases the postulated rupture location. The sleeve contains circumferential grooves on the inside surface to accommodate collars that are welded to the pipe and to hold the ISR in place after its installation and during plant operations. In the event of a pipe rupture, the collars are captured by the ISR grooves, thus preventing the pipe from whipping. The ISR, which encloses the rupture, also restricts the escaping fluid and shields the surroundings from fluid jets. A typical ISR is shown in Figure 3.6-33 and Figure 3.6-34. The ISRs are designed to be compatible with the plant design so that when an ISR is required at a specific location, the impacts to the design of the piping system and to plant operations and maintenance are minimized. Because ISRs fit closely around the pipe, the physical envelope of the ISR is small and unlikely to interfere with neighboring components. Additionally, because an ISR is fully supported by the encased piping, additional supporting structures are not necessary. The only requirements for the implementation of an ISR at a particular location are that there is sufficient space on either side of the postulated break that is free of interferences such that the ISR can be designed and installed, and collars can be placed. The ISR is designed to be removable for inspection of the piping welds if required, with sufficient clearance between the collars and welds to provide for ultrasonic inspection of each weld. To achieve this, the ISR sleeve is fabricated in two halves and is bolted in place over the postulated break location. The methodology used to size and qualify the ISRs is described below. In order for the ISR to perform its function without failing, the design considers the jet thrust load, which pushes the broken pipe ends apart, the internal pressure increase acting on the sleeve, as well as the preexisting loads that were carried by the piping prior to rupture. To compute the static jet thrust load due to a jet from a broken pipe end, the guidance in SRP 3.6.2 Revision 2, Section III.2.C. (iv) is employed. This is conservative because no actual 2 3.6-44 Revision 1

using the simplified approach described in SRP 3.6.2, as T = (K)(P)(A) Eq. 3.6-53 T = jet thrust force, pounds K = thrust coefficient equal to 1.26 for steam, saturated water or steam-water mixtures, 2.0 for subcooled, non-flashing water P = system pressure prior to the pipe break, (psi) A = pipe break area, (in2) The thrust due to pipe break is computed with the full area of the pipe, including the metal area, because of the confinement of the sleeve, even though the sleeve includes holes for pressure relief. The system pressures used to compute the pipe break thrust load are assumed to be the normal operating pressure of the piping. System pressure for the CVCS is 1850 psia. The design of the ISRs is such that thrust loads are taken in pure shear by the collars and sleeve with bearing on the collars and sleeve where contact occurs. This design method, therefore, limits the primary stresses on individual critical parts of the ISR. The thickness of each collar is less than or equal to the wall thickness of the pipe to which it is attached, for welding considerations, and a full penetration weld with stress-relief treatment is used. The thickness of the collars are specified to meet the allowable shear stress criteria. To preclude pullout of the collar, the collar long axis is designed to be perpendicular to the pipe longitudinal axis. The collars are completely captured by the sleeve grooves. The sleeve grooves are designed to receive the collars, ensuring the shear forces are applied near the base of the collars to minimize bending moment on the collars. To accommodate the pressure increase internal to the ISR following a pipe rupture, the design uses a pressure relief chamber. The pressure relief chamber is shown in Figure 3.6-34. The chamber includes holes through which fluid from the postulated pipe break is released, thereby relieving pressure inside the ISR. However, for qualification, the pressure acting on the onside surface of the chamber is conservatively assumed to be the initial operating pressure in the pipe prior to the break. The essential features of the pressure relief chamber are:

the same outside diameter as that of the sleeve.
an increased inside diameter to permit the chamber to be formed.
pressure relief holes to permit fluid to be released in a controlled way.
raised longitudinal ridges on the inside surface of the pressure relief chamber maintain a consistent gap with the outside diameter of the pipe along the entire length of the ISR
The holes are arranged symmetrically on each side of the postulated pipe break weld.

2 3.6-45 Revision 1

straight pipe to straight pipe configuration because of the following considerations.

Breaks at straight pipe connections to a long neck flange are postulated; custom flanges may be used which are similar to a straight pipe to straight pipe configuration.
Safe end to straight pipe welds may be made the same as straight pipe to straight pipe welds by use of long safe ends.

Structural evaluation of the ISR design is performed using finite element analysis. The finite element code ANSYS is used to perform both a linear and a nonlinear analysis. A linear analysis is used to qualify the ISR to ASME Level D requirements, while a separate plastic analysis is performed to verify that the piping segments will collapse before if the ISR experiences excessive plasticity. Both analyses utilize a 3D finite element model which includes the sleeve, eight bolts, washers and nuts, and two piping sections with collars, which represent the broken pipe which is being restrained. Linear Analysis The ISR is analyzed to Level D limits specified within Appendix F of the ASME Boiler and Pressure Vessel Code. The ISR and the pipe collar are designed using the limits given in F-1331. The ISR bolts are designed using the limits given in F-1335. Loads included in the analysis are: an overturning moment equal to the maximum moment carried by the pipe applied to a pipe segment, thrust forces applied to each pipe segment, pressure applied the cavity between the ISR and the pipe, and a preload added to each bolt. Cut lines in the model are used to analyze the stresses in both the ISR and the collar. It should be noted that this is a conservative approach to calculate the membrane stress. Membrane stress is defined as normal stress that is uniformly distributed and equal to the average stress across the thickness of the section under consideration. The stresses extracted by the cut lines are stresses averaged over a single line at a highly stressed location, not a whole section, and therefore produce conservative results. The results of the linear finite element analysis of the ISR are provided in the tables below. ISR Stress Results Cut Line Maximum Stress Classification Allowable (psi) Ratio Stress (psi) Membrane 38335 38880 0.99 Membrane Plus Bending 55515 58320 0.95 Bolt Stress Results Check Stress (psi) Allowable (psi) Ratio Tension 56661 94430 0.60 Tension + Bending 91997 134900 0.68 2 3.6-46 Revision 1

Check Stress (psi) Allowable (psi) Ratio Average Shear Stress 8787 26628 0.33 Plastic Analysis A plastic analysis is performed to verify that the piping will collapse before the ISR. This analysis is not a code requirement and is only done to determine if the ISR itself can withstand a greater moment than the piping system. The model used in the linear analysis is also used for this analysis, however is modified to increase the length of the piping segments. All other geometry is kept the same. The boundary conditions are identical to the linear analysis with the exception of the applied moment. The maximum moment was increased to ensure a moment high enough to cause the piping system to collapse was applied. For the material properties, the bolts use bi-linear kinematic hardening curve while the ISR and piping segments use a multilinear kinematic hardening curve. The last four inches of the piping segments use elastic material properties in order to help the model converge. The analysis shows that the pipe and collar have developed through wall plastic strains prior to the ISR developing through wall plastic strains. Therefore the piping segments will collapse prior to the ISR. The criteria for postulating ruptures, and thus locating ISRs, are discussed in Section 3.6.2 and Section 3.6.4. If stress criteria are used (as opposed to assuming a rupture at every weld and fitting), then these criteria are evaluated during the code stress analysis of the piping systems. The design of the ISRs is such that the impacts to this analysis for piping systems that use ISRs are minimized. If it is determined during the analysis that an ISR is required at a location, the weight (lumped mass) of the ISR is added to the piping model and the analysis is performed again. This process is iterated until the piping passes its code analysis criteria while accounting for the added mass of the required ISRs. The ISRs are designed and located with sufficient clearance between the pipe and the ISR such that they do not normally interact and cause additional piping stresses. A design hot position gap is provided, which allows maximum predicted displacements (e.g., thermal and seismic) to occur without ISR interaction. A total of 27 ISRs are employed inside the containment vessel. This includes 17 on the RCS injection, discharge, high point vent, and pressurizer spray lines, and two on the DHRS condensate return lines. In the reactor pool bay, there are an additional 12 ISRs: each CVCS line (RCS injection, discharge, high point vent, and pressurizer spray line) has three. In addition to the ISRs directly associated with the NPM, additional ISRs may be used within the RXB, depending on the routing of COL applicant scope high- and moderate-energy piping beyond the NPMs and through the BOP. 2 3.6-47 Revision 1

The ISR shown in Figure 3.6-33 and Figure 3.6-34, and whose sizing methodology is discussed above, is intended to mitigate the adverse effects of a circumferential pipe break. Although longitudinal pipe breaks may also be required to be postulated per the criteria given in BTP 3-4, there are currently none being postulated. An objective of the piping design is to design the piping such that stress and fatigue limits as specified in Section 3.6.2 are satisfied, thus precluding the need to postulate pipe breaks in the run piping (i.e. locations other than terminal ends). Therefore, the majority of postulated breaks are located at terminal ends. Per the criteria in BTP 3-4, longitudinal pipe breaks need not be postulated at terminal ends. Current detailed stress analyses of piping systems have identified only two locations where piping ruptures must be postulated in locations other than terminal ends, both of which are located on NPS 2 piping. Additional criteria in BTP 3-4 excludes the postulation of longitudinal breaks in piping of sizes less than NPS 4. The ISR designs will continue to be developed to be compatible with the various configurations of piping where breaks are postulated and with the types of break which must be postulated, including longitudinal breaks, as the detailed piping analyses indicate they are required. 5.1 Integral Jet Impingement Shield and Pipe Whip Restraint Computational Fluid Dynamics Analysis The dynamic jet effect for a postulated pipe break was evaluated as required by GDC 4 and SRP 3.6.2. Computational fluid dynamics (CFD) models were developed of the flow exiting the ISR following the pipe break. The CFD models were used to determine the loads on nearby components caused by jet impingement. ANSYS CFX was used for the detailed CFD evaluation, utilizing an axisymmetric 2D model to capture the flow out of a single ISR hole. The RCS pipe break case uses a high temperature of 550 degrees F that bounds the cold leg temperatures. At 550 degrees F the saturation pressure is 1,045 psia. The ISR relief hole exit flow is taken at a pressure of 1,044 psia, a temperature of 550 degrees F, and a constant inlet speed of 1,602 ft/s (the speed of sound at these conditions). The DHRS has both a high temperature condition that flashes to steam and a low temperature condition that does not flash. The DHRS high temperature pipe break uses a temperature of 310 degrees F, an inlet pressure of 77 psia, and a constant inlet speed of 1,624 ft/s. The DHRS low temperature pipe break uses a temperature of 40 degrees F and an inlet pressure of 800 psia. 800 psia correlates to an inlet speed of 341 ft/s. The CFD models are 5 degree wedges with side planes set to symmetry boundary conditions. The ISR walls are set to be free-slip walls. The inlet is set to supersonic for the RCS break and DHRS high temperature break cases and subsonic for the DHRS low temperature case. For the RCS ISR, Figure 3.6-36 shows a graph of total pressure along the axis. Note that X=0 is the ISR relief hole exit with +X normal to the hole. The total pressure drops below 50 psi within the first three inches. Figure 3.6-37 shows a plot of total pressure above the axis at distances of 5, 10, 15, and 20 inches away from the relief hole exit, with Y=0 2 3.6-48 Revision 1

gives the radius of the circular jet area). Total pressure is reported as relative pressure. For absolute total pressure, add 1 atm. Figure 3.6-38 shows a graph of total pressure along the axis for the DHRS high temperature ISR case. Note that X=0 is at the ISR relief hole exit with +X going normal to the hole. The total pressure drops below 20 psi within the first five inches. For the DHRS low temperature ISR case, Figure 3.6-39 shows a graph of total pressure along the discharge access. Note that X=0 is at the ISR relief hole exit with +X normal to the hole. Figure 3.6-40 shows a plot of total pressure above the axis at distances of 5, 10, 15, and 20 inches away from the relief hole exit, with Y=0 at the centerline. The plots above the axis demonstrate how large the jet is at these distances (force is calculated by multiplying total pressure and area, where Figure 3.6-40 provides the radius of the circular jet area). Total pressure is relative pressure, for absolute total pressure, add 1 atm. As demonstrated by these CFD analysis results, for the RC,S ISR, and the DHRS high temperature ISR applications, total pipe break discharge pressure within five inches of the ISR drops below 50 psia. The lower temperature DHRS ISR application has higher total pressures that extend farther radially from the pipe break, due to the absence of discharge flashing. Application of the NuScale ISR device in areas with common pipe routing and essential SSC spacing therefore assures the mitigation of detrimental jet impingement effects to nearby safety-related, risk significant SSC. Jet impingement loads at reasonable radial distances from the ISR are low, allowing for proper design and placement of vicinity SSC. 5.2 Integral Jet Impingement Shield and Pipe Whip Restraint Confirmatory Test Program As the NuScale ISR is a first-of-a-kind jet impingement shield and pipe whip restraint, proof of concept testing is being performed to validate the analytical model and demonstrate that the ISR performs its intended function to mitigate the dynamic effects of postulated high energy pipe breaks. The ISR test objectives are to measure the total pressure in the jet exiting the ISR as a function of distance from the ISR, to measure the pressure inside the ISR chamber to validate the analytical models, to measure the acceleration on the pipe ends during a simulated pipe break to validate the structural design, and to confirm the ability to fabricate and install the ISR on a prototypic section of pipe. The test facility replicates the transient nature of a pipe break by having two ends of pipe initially held together. After the simulated break is initiated, the two ends of pipe accelerate in opposite directions and a gap is established between the two separated ends of pipe. The gap opening is controlled by the clearance between the ISR grooves and the welded collars on the pipe. Fluid escapes the pipe through the gap into the ISR 2 3.6-49 Revision 1

Of the high-energy piping systems that use ISRs, the CVCS lines contain the highest pressure. The CVCS injection and discharge lines are approximately at RCS pressure and contain subcooled liquid water. The temperature of the discharge line is approximately equal to the RCS downcomer temperature. The as-installed clearance between the pipe collars and the ISR grooves is measured before each test. The gap length between the ends of the pipe is measured after each test so that the break area can be calculated. The ISR is disassembled and inspected after each test to ensure that the dimensions meet the fabrication drawings and components performed as intended. The facility design and closure force minimize leakage between the pipe ends before the simulated break in order that leakage not accumulate in the ISR chamber during startup. The opening time for the ISR gap is as small as achievable to replicate a realistic pipe break. Pressure instrumentation is provided inside the chamber of the ISR. Additional pressure instruments are located outside the ISR to measure the pressure distribution as a function of distance from the ISR. Recorded test parameters include pipe internal pressure and temperature at the break, pipe flow rate through the break, ISR chamber pressure, pressure external to the ISR, and pipe acceleration. 6 References 3.6-1 Electric Power Research Institute, "An Engineering Approach for Elastic-Plastic Fracture Analysis," NP-1931, July 1981. 3.6-2 Electric Power Research Institute, "Advances in Elastic-Plastic Fracture Analysis," NP-3607, August 1984. 3.6-3 Electric Power Research Institute, "Elastic-Plastic Fracture Analysis of Through-Wall and Surface Flaws in Cylinders," NP-5596, January 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 U.S. Nuclear Regulatory Commission, "The Development of a J-estimation Scheme for Circumferential and Axial Through-wall Cracked Elbows," NUREG/ CR-6837, Vol. 2, Appendix F, June 2005. 3.6-6 Electric Power Research Institute, "Crack-Opening Area Calculations for Circumferential Through-Wall Pipe Cracks," NP-5959-SR, August 1988. 2 3.6-50 Revision 1

3.6-8 Henry, R.E., "The Two-Phase Critical Discharge of Initially Saturated or Subcooled Liquid," Nuclear Science and Engineering, Vol. 41, pp. 336-342, 1970. 3.6-9 Henry, R. E. and Fauske, H. K., "Two-Phase Critical Flow at Low Qualities, Part I: Experimental," Nuclear Science and Engineering, Vol. 41, pp. 79-91, 1970. 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 NuScale Power, LLC, Reactor Module Test and Inspection Elements, ER-A010-2186, Revision 0. 3.6-12 ASME Operation and Maintenance Code Assessment, ER-A010-3875, Revision 0. 3.6-13 U.S. Nuclear Regulatory Commission, "Guidance on Monitoring and Responding to Reactor Coolant System Leakage," Regulatory Guide 1.45, Revision 1, May 2008. 3.6-14 "Improved J and COD Estimation by GE/EPRI Method in Elastic to Fully Plastic Transition Zone," Engineering Fracture Mechanics, Volume 73, Issue 14, pages 1959-1979 September 2006. 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. 2 3.6-51 Revision 1

Table 3.6-1: High- and Moderate-Energy Fluid System Piping 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 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 3.6-52 Revision 1

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 High 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 3.6-53 Revision 1

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 nozzle-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 3.6-54 Revision 1

Table 3.6-2: Postulated Break Locations Line ASME Class Postulated Break Location (see Figure 3.6-6 thru Figure 3.6-11) k locations inside containment RCS injection 1 Terminal end - RPV head (Figure 3.6-6) Tee/pipe weld Pipe-to-pipe weld Valve/pipe weld Terminal end - containment boundary RCS discharge 1 Terminal end - RPV head (Figure 3.6-7) Valve/pipe weld Pipe-to-pipe weld Terminal end - containment boundary Pressurizer spray 1 Terminal end - RPV head (Figure 3.6-8) Valve/pipe weld Terminal end - containment boundary Tee/pipe weld RCS high-point vent 1 Terminal end - RPV head (Figure 3.6-9) Valve/pipe weld Pipe-to-pipe weld Terminal end - containment boundary DHRS #1 2 Terminal end - containment boundary (Figure 3.6-10) DHRS #2 2 Terminal end - containment boundary (Figure 3.6-11) k locations outside the CNV to the NPM disconnect flange RCS injection 3 Valve/pipe weld (Figure 3.6-12) Tee/pipe weld Tee/flange weld RCS discharge 3 Valve/pipe weld (Figure 3.6-12) Tee/pipe weld Tee/flange weld Pressurizer spray 3 Valve/pipe weld (Figure 3.6-13) Tee/pipe weld Tee/flange weld RCS high-point vent 3 Valve/pipe weld (Figure 3.6-13) Tee/pipe weld Tee/flange weld k locations in the RXB (outside the NPM disconnect flange), documented in the NuScale Pipe Rupture Hazards Analysis. 2 3.6-55 Revision 1

Table 3.6-3a: Summary of Main Steam Line Bounding Analysis Curves NPS 8 Welds (Pipe- to-PS 8 Base Metal Pipe and Pipe-to-Safe- NPS 12 Base Metal NPS 12 Welds NPS 8 Elbow End) rmal Max Stress Normal Max Stress Normal Max Stress Normal Max Stress Normal Max Stress ss (psi) (psi) Stress (psi) (psi) Stress (psi) (psi) Stress (psi) (psi) Stress (psi) (psi) ,136 4,571 3,136 5,334 1,229 2,383 1,229 3,089 3,136 3,819 ,136 10,741 5,136 11,447 1,309 2,753 1,309 3,517 5,136 8,910 ,136 23,204 9,136 22,114 3,229 10,826 3,229 12,613 9,136 18,896

,136     33,733    13,136      31,290      5,229       17,876    5,229     19,807    18896       18,896
,136     41,056    17,136      38,738      9,229       30,077    9,229     30,789
,136     45,548    21,136      44,580     13,229       39,046   13,229     39,078
,136     47,720    25,136      49,026     17,229       44,763   17,229     45,324
,136     49,131    29,136      52,380     21,229       47,966   21,229     49,915 25,229       49,526   25,229     53,274 29,229       50,534   29,229     55,715 2                                                3.6-56                                        Revision 1

Table 3.6-3b: Summary of Feedwater System Line Bounding Analysis Curves NPS 4 Base Metal NPS 4 Welds NPS 5 Base Metal NPS 5 Welds ormal Max Stress Normal Max Stress Normal Max Stress Normal Stress Max Stress ress psi psi Stress psi psi Stress psi psi psi psi 2,052 6,698 969 1,100 1,079 4,611 1,079 5,465 3,135 11,722 2,200 8,648 2,160 11,692 2,308 14,294 4,219 15,872 3,430 14,718 3,241 17,021 3,536 20,677 5,302 19,419 4,661 19,592 4,321 21,218 4,764 25,563 7,469 25,358 5,892 23,666 5,402 24,691 5,992 29,516 9,635 30,227 8,354 30,318 7,563 30,233 8,448 35,615 11,802 34,232 10,815 35,553 9,724 34,548 10,904 40,302 13,968 37,641 13,277 39,878 11,885 38,104 13,360 44,110 16,135 40,516 15,738 43,529 14,046 41,058 15,816 47,253 18,301 42,925 18,200 46,600 16,207 43,510 18,272 49,860 20,468 44,936 20,661 49,175 18,368 45,543 20,728 52,025 22,634 46,616 23,123 51,330 20,529 47,229 23,184 53,827 26,968 49,204 25,584 53,136 22,690 48,631 25,640 55,331 31,301 51,052 30,508 55,931 27,012 50,784 30,552 57,653 44,300 54,180 35,431 57,938 31,334 52,317 35,464 59,319 50,200 61,361 44,300 54,908 50,200 62,153 2 3.6-57 Revision 1

cess System ASME NPS Design Operating 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 1000 550 80 121 (CNTS) CFDS NTS-inside Class 2 2 165 300 85 100 CNV) CFDS TS-outside Class 2 4 1000 550 85 100 CNV) CES Class 2 4 1000 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. 2 3.6-58 Revision 1

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 3.6-59 Revision 1

Table 3.6-6: Allowable Stresses for Class 1 Piping (ksi) Process System 2.4Sm 2.25Sm 1.8Sy 1.2Sm CVCS (RCS) 38.88 36.45 32.40 19.44 2 3.6-60 Revision 1

Table 3.6-7: Allowable Stresses for Class 2 & 3 Piping (ksi) 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 3.6-61 Revision 1

Table 3.6-8: Jet loads and Maximum Bending Moments Pipe Jet 5D Bending Plastic R=M/ cess System Size Load Bend Moment, M Moment, Mp Mp (NPS) (kip) (in.) (in-kip) (in-kip) CVCS 2 8.36 10 83.6 26.07 3.21 4 11.35 20 226.9 162.51 1.40 FWS 5 17.99 25 449.7 288.03 1.56 2 6.26 10 62.6 25.93 2.41 DHRS 6 59.17 30 1775.0 456.18 3.89 8 25.56 40 1022.3 838.21 1.22 MSS 12 57.18 60 3430.8 2574.16 1.33 2 3.6-62 Revision 1

Protection against Dynamic Effects Associated with Postulated Rupture cale Final Safety Analysis Report of Piping

Figure 3.6-2: Main Steam Line 1 CONTAINMENT BOUNDARY NPS 12 NPS 8 TO STEAM PLENUM SGS MAIN STEAM LINE #1 HIGH-ENERGY INSIDE CONTAINMENT 8" AND 12" NOMINAL DIAMETER, NO HORIZONTAL RUNS NO BREAKS POSTULATED (QUALIFIES AS LBB). 2 3.6-64 Revision 1

Figure 3.6-3: Main Steam Line 2 CONTAINMENT BOUNDARY NPS 12 NPS 8 TO STEAM PLENUM SGS MAIN STEAM LINE #2 HIGH-ENERGY INSIDE CONTAINMENT 8" AND 12" NOMINAL DIAMETER, NO SIGNIFICANT HORIZONTAL RUNS NO BREAKS POSTULATED (QUALIFIES AS LBB). 2 3.6-65 Revision 1

Figure 3.6-4: Feedwater Line 1 CONTAINMENT BOUNDARY NPS 5 DHRS PIPING (REF) NPS 4 NPS 4 TO FEEDWATER PLENUM FEEDWATER LINE 1 HIGH-ENERGY INSIDE CONTAINMENT 4" - 5" NOMINAL DIAMETER, NO HORIZONTAL RUNS NO BREAKS POSTULATED (QUALIFIES AS LBB) 2 3.6-66 Revision 1

Figure 3.6-5: Feedwater Line 2 CONTAINMENT BOUNDARY NPS 5 DHRS PIPING (REF) NPS 4 NPS 4 TO FEEDWATER PLENUM FEEDWATER LINE 2 HIGH-ENERGY INSIDE CONTAINMENT 4" - 5" NOMINAL DIAMETER, NO HORIZONTAL RUNS NO BREAKS POSTULATED (QUALIFIES AS LBB) 2 3.6-67 Revision 1

igure 3.6-6: Chemical and Volume Control System - Reactor Coolant System Injection Line Postulated Break Locations CONTAINMENT BOUNDARY POSTULATED BREAK LOCATION NPS 2 POSTULATED BREAK LOCATION B DETAIL B NPS 2 POSTULATED BREAK LOCATION CHECK VALVE A POSTULATED BREAK LOCATION POSTULATED BREAK LOCATION RPV VESSEL HEAD DETAIL A CVC SYSTEM RCS INJECTION LINE HIGH-ENERGY INSIDE CONTAINMENT 2" NOMINAL DIAMETER BREAKS POSTULATED AT TERMINAL ENDS, VALVE WELD, AND TEE WELDS. 2 3.6-68 Revision 1

igure 3.6-7: Chemical and Volume Control System - Reactor Coolant System Discharge Line Postulated Break Locations CONTAINMENT BOUNDARY POSTULATED BREAK LOCATION NPS 2 POSTULATED BREAK LOCATION CHECK VALVE POSTULATED BREAK LOCATION RPV VESSEL HEAD CVC SYSTEM RCS DISCHARGE LINE HIGH-ENERGY INSIDE CONTAINMENT 2" NOMINAL DIAMETER BREAKS POSTULATED AT TERMINAL ENDS, VALVE WELD. 2 3.6-69 Revision 1

gure 3.6-8: Chemical and Volume Control System - Pressurizer Spray Line Postulated Break Locations CONTAINMENT BOUNDARY POSTULATED BREAK LOCATION NPS 2 POSTULATED BREAK LOCATION POSTULATED BREAK LOCATION NPS 2 POSTULATED BREAK LOCATION POSTULATED BREAK LOCATION CHECK VALVE CHECK VALVE RPV VESSEL HEAD CVC SYSTEM PZR SPRAY SUPPLY LINE HIGH-ENERGY INSIDE CONTAINMENT 2" NOMINAL DIAMETER BREAKS POSTULATED AT TERMINAL ENDS, VALVE WELD, AND TEE WELDS. 2 3.6-70 Revision 1

ure 3.6-9: Chemical and Volume Control System - High Point Vent Postulated Break Locations CONTAINMENT BOUNDARY POSTULATED BREAK LOCATION NPS 2 POSTULATED BREAK LOCATION POSTULATED BREAK LOCATION RPV VESSEL HEAD CHECK VALVE CVC SYSTEM HIGH POINT DEGASIFICATION LINE HIGH-ENERGY INSIDE CONTAINMENT 2" NOMINAL DIAMETER BREAKS POSTULATED AT TERMINAL ENDS, VALVE WELD, AND PIPE-TO-PIPE WELD. 2 3.6-71 Revision 1

Figure 3.6-10: Decay Heat Removal System Line 1 Postulated Break Locations MS PIPING (REF) ISOLATION VALVE ISOLATION VALVE DHRS SUPPORT NPS 6 DHRS LINE OUTSIDE CONTAINMENT DHRS PASSIVE CONDENSER (REF) FEEDWATER PIPE (REF) DHRS LINE INSIDE CONTAINMENT DHRS LINE OUTSIDE CONTAINMENT POSTULATED BREAK LOCATION NPS 2 (CONTAINMENT WALL) DECAY HEAT REMOVAL SYSTEM LINE #1 HIGH-ENERGY INSIDE AND OUTSIDE CONTAINMENT 2" NOMINAL DIAMETER INSIDE CONTAINMENT BREAKS (INSIDE CONTAINMENT) POSTULATED AT LOCATIONS INDICATED. 2" AND 6" NOMINAL DIAMETER OUTSIDE CONTAINMENT (NO BREAKS POSTULATED OUTSIDE CONTAINMENT, PIPING QUALIFIES TO BTP 3-4 B.A. (ii)). 2 3.6-72 Revision 1

Figure 3.6-11: Decay Heat Removal System Line 2 Postulated Break Locations MS PIPING (REF) ISOLATION VALVE ISOLATION VALVE DHRS SUPPORT NPS 6 DHRS LINE OUTSIDE CONTAINMENT DHRS PASSIVE CONDENSER (REF) FW PIPE (REF) DHRS LINE INSIDE CONTAINMENT NPS 2 POSTULATED BREAK LOCATION DHRS LINE OUTSIDE CONTAINMENT (CONTAINMENT WALL) DECAY HEAT REMOVAL SYSTEM LINE #2 HIGH-ENERGY INSIDE AND OUTSIDE CONTAINMENT 2" NOMINAL DIAMETER INSIDE CONTAINMENT BREAKS (INSIDE CONTAINMENT) POSTULATED AT LOCATIONS INDICATED. 2" AND 6" NOMINAL DIAMETER OUTSIDE CONTAINMENT (NO BREAKS POSTULATED OUTSIDE CONTAINMENT, PIPING QUALIFIES TO BTP 3-4 B.A. (ii)). 2 3.6-73 Revision 1

ure 3.6-12: Containment System Chemical and Volume Control Discharge and Injection Line Postulated Break Locations CNTS PIPING BOUNDARY POSTULATED BREAK LOCATION NPS 2 POSTULATED BREAK LOCATION DUAL CONTAINMENT ISOLATION VALVE CNTS CVC DISCHARGE LINE OUTSIDE CONTAINMENT 2" NOMINAL DIAMETER POSTULATED BREAKS INDICATED CNTS PIPING BOUNDARY POSTULATED BREAK LOCATION NPS 2 POSTULATED BREAK LOCATION DUAL CONTAINMENT ISOLATI2N VALVE CNTS CVC INJECTION LINE OUTSIDE CONTAINMENT 2" NOMINAL DIAMETER POSTULATED BREAKS INDICATED 2 3.6-74 Revision 1

Figure 3.6-13: Chemical and Volume Control System Postulated Break Locations CNTS PIPING BOUNDARY POSTULATED BREAK LOCATION NPS 2 DUAL CONTAINMENT ISOLATION VALVE POSTULATED BREAK LOCATION CNTS CVC PRESSURIZER SPRAY LINE OUTSIDE CONTAINMENT 2" NOMINAL DIAMETER POSTULATED BREAKS INDICATED CNTS PIPING BOUNDARY POSTULATED BREAK LOCATION NPS 2 POSTULATED BREAK LOCATION DUAL CONTAINMENT ISOLATION VALVE CNTS RPV HIGH POINT DEGASIFICATION LINE OUTSIDE CONTAINMENT 2" NOMINAL DIAMETER POSTULATED BREAKS INDICATED 2 3.6-75 Revision 1

Figure 3.6-14: Feedwater Line Postulated Break Locations CNTS PIPING BOUNDARY NPS 4 FEEDWATER ISOLATION VALVE CNTS FEEDWATER LINE #1 CNTS PIPING BOUNDARY NPS 4 FEEDWATER ISOLATION VALVE CNTS FEEDWATER LINE #2 CNTS FEEDWATER LINE #1 AND LINE #2 OUTSIDE CONTAINMENT 4" NOMINAL DIAMETER 12%5($.632678/$7('2876,'(&217$,10(17 3,3,1*48$/,),(672%73%$ LL 2 3.6-76 Revision 1

Figure 3.6-15: Main Steam Line Postulated Break Locations CNTS PIPING BOUNDARY NPS 12 MAIN STEAM ISOLATION VALVE NPS 12 CNTS MAIN STEAM LINE #1 CNTS PIPING BOUNDARY NPS 12 MAIN STEAM ISOLATION VALVE NPS 12 CNTS MAIN STEAM LINE #2 HIGH ENERGY OUTSIDE CONTAINMENT 12" NOMINAL DIAMETER 12%5($.632678/$7('2876,'(&217$,10(17 3,3,1*48$/,),(672%73%$ LL 2 3.6-77 Revision 1

Protection against Dynamic Effects Associated with Postulated Rupture cale Final Safety Analysis Report of Piping Figure 3.6-16: Postulated High-Energy Main Steam System Pipe Routing Beyond the NuScale Power Module (COL Applicant Scope)

Protection against Dynamic Effects Associated with Postulated Rupture cale Final Safety Analysis Report of Piping Figure 3.6-17: Postulated High-Energy Feedwater System Pipe Routing Beyond the NuScale Power Module (COL Applicant Scope)

Figure 3.6-18: Flow Chart for Piping Leak-Before-Break Evaluation 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 3.6-80 Revision 1

Figure 3.6-19: Illustration of Pipe with a Circumferential Through-Wall Crack 2 3.6-81 Revision 1

Figure 3.6-20: Henry-Fauske's Model of Two-Phase Flow 2 3.6-82 Revision 1

Figure 3.6-21: Local and Global Surface Roughness and Turns 2 3.6-83 Revision 1

Figure 3.6-22: Crack Opening Displacement-Dependent Effective Crack Morphology 2 3.6-84 Revision 1

igure 3.6-23: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Straight Pipe Base Metal 8" MS Line Straight Base Metal 60,000 50,000 40,000 30,000 Min. LR = 0.20 lbm/min 20,000 MS Line 1 Stress Points MS Line 2 Stress Points 10,000 0 0 5,000 10,000 15,000 20,000 25,000 30,000 Normal Stress, psi 2 3.6-85 Revision 1

igure 3.6-24: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Pipe-to-Pipe Weld 8" MS Line Pipe-to-Pipe Weld 60,000 50,000 40,000 30,000 Min. LR = 0.20 lbm/min 20,000 MS Line 1 Stress Points MS Line 2 Stress Points 10,000 0 0 5,000 10,000 15,000 20,000 25,000 30,000 Normal Stress, psi 2 3.6-86 Revision 1

igure 3.6-25: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Pipe-to-Safe-End Weld 8" MS Line Pipe-to-Safe End Weld 60,000 50,000 40,000 30,000 Min. LR = 0.20 lbm/min 20,000 MS Line 1 Stress Points MS Line 2 Stress Points 10,000 0 0 5,000 10,000 15,000 20,000 25,000 30,000 Normal Stress, psi 2 3.6-87 Revision 1

gure 3.6-26: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 12 Straight Pipe Base Metal 12" MS Line Straight Base Metal 60,000 50,000 40,000 30,000 Min. LR = 0.20 lbm/min 20,000 MS Line 1 Stress Points MS Line 2 Stress Points 10,000 0 0 5,000 10,000 15,000 20,000 25,000 30,000 Normal Stress, psi 2 3.6-88 Revision 1

gure 3.6-27: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 12 Pipe-to-Safe-End Weld 12" MS Line Pipe-to-Safe-End Weld 60,000 50,000 40,000 30,000 Min. LR = 0.20 lbm/min 20,000 MS Line 1 Stress Points MS Line 2 Stress Points 10,000 0 0 5,000 10,000 15,000 20,000 25,000 30,000 Normal Stress, psi 2 3.6-89 Revision 1

igure 3.6-28: Smooth Bounding Analysis Curve for Main Steam System Nominal Pipe Size 8 Elbow Base Metal 8" MS Line Elbow Base Metal 20,000 15,000 10,000 Min. LR = 0.20 lbm/min MS Line 1 Stress Points 5,000 MS Line 2 Stress Points 0 0 5,000 10,000 15,000 20,000 25,000 30,000 Normal Stress, psi 2 3.6-90 Revision 1

ure 3.6-29: Smooth Bounding Analysis Curve for Nominal Pipe Size 4 Feedwater System Line Base Metal 4" FW Line Base Metal 60,000 50,000 40,000 30,000 20,000 Min. LR = 0.20 lbm/min 10,000 FW Line 1 Stress Points FW Line 2 Stress Points 0 0 5,000 10,000 15,000 20,000 25,000 30,000 Normal Stress, psi 2 3.6-91 Revision 1

ure 3.6-30: Smooth Bounding Analysis Curve for Nominal Pipe Size 4 Feedwater System Line Welds 4" FW Line Welds 70,000 60,000 Maximum Stress, psi 50,000 40,000 30,000 20,000 Min. LR = 0.20 lbm/min 10,000 FW Line 1 Stress Points FW Line 2 Stress Points 0 0 5,000 10,000 15,000 20,000 25,000 30,000 Normal Stress, psi 2 3.6-92 Revision 1

ure 3.6-31: Smooth Bounding Analysis Curve for Nominal Pipe Size 5 Feedwater System Line Base Metal 5" FW Line Base Metal 70,000 60,000 50,000 40,000 30,000 20,000 Min. LR = 0.20 lbm/min 10,000 FW Line 1 Stress Points FW Line 2 Stress Points 0 0 5,000 10,000 15,000 20,000 25,000 30,000 Normal Stress, psi 2 3.6-93 Revision 1

ure 3.6-32: Smooth Bounding Analysis Curve for Nominal Pipe Size 5 Feedwater System Line Welds 5" FW Line Welds 70,000 60,000 50,000 40,000 30,000 20,000 Min. LR = 0.20 lbm/min 10,000 FW Line 1 Stress Points FW Line 2 Stress Points 0 0 5,000 10,000 15,000 20,000 25,000 30,000 Normal Stress, psi 2 3.6-94 Revision 1

Protection against Dynamic Effects Associated with Postulated Rupture cale Final Safety Analysis Report of Piping Figure 3.6-33: Typical Integral Jet Impingement Shield and Pipe Whip Restraint A010.704 A010.703 A010.705 A010.701

Protection against Dynamic Effects Associated with Postulated Rupture cale Final Safety Analysis Report of Piping Figure 3.6-34: Cutaway View of Integral Jet Impingement Shield and Pipe Whip Restraint

Figure 3.6-35: Disk-Type Jet from Circumferential Pipe Rupture at a Weld 2 3.6-97 Revision 1

Figure 3.6-36: RCS pipe break total pressure drop along discharge centerline 2 3.6-98 Revision 1

gure 3.6-37: RCS pipe break total pressure graph at 5, 10, 15, and 20 inches radially from ISR 2 3.6-99 Revision 1

Figure 3.6-38: DHRS high temperature pipe break total pressure graph along discharge centerline 2 3.6-100 Revision 1

ure 3.6-39: DHRS low temperature pipe break total pressure drop along discharge centerline 2 3.6-101 Revision 1

ure 3.6-40: DHRS low temperature pipe break total pressure graph at 5, 10, 15, and 20 inches radially from ISR 2 3.6-102 Revision 1}}

ML18086A153

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