Text

Part 02 - Final Safety Analysis Report (Rev. 1) - Part 02 - Tier 02 - Chapter 14 - Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria - Sections 14.00 - 14.03
	ML18086A185
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: ML18086A185 (314)
	v • d • e

NuScale Standard Plant Design Certification Application Chapter Fourteen Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria PART 2 - TIER 2 Revision 1 March 2018

COPYRIGHT NOTICE This document bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this document, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.

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

NuScale Final Safety Analysis Report Table of Contents TABLE OF CONTENTS CHAPTER 14 INITIAL TEST PROGRAM AND INSPECTIONS, TESTS, ANALYSES, AND ACCEPTANCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.0-1 14.0 Verification Programs. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.0-1 14.1 Specific Information to be Addressed for the Initial Plant Test Program . . . . . . . 14.1-1 14.2 Initial Plant Test Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-1 14.2.1 Summary of Initial Test Program and Objectives . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-1 14.2.2 Organization and Staffing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-6 14.2.3 Test Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-6 14.2.4 Conduct of the Test Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-11 14.2.5 Review, Evaluation, and Approval of Test Results. . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-11 14.2.6 Test Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-12 14.2.7 Test Programs Conformance with Regulatory Guides . . . . . . . . . . . . . . . . . . . . . . 14.2-12 14.2.8 Utilization of Reactor Operating and Testing Experience in Test Program Development. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-13 14.2.9 Trial Use of Plant Operating Procedures, Emergency Procedures, and Surveillance Procedures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-13 14.2.10 Initial Fuel Loading, and Initial Criticality. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-14 14.2.11 Test Program Schedule and Sequence . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-17 14.2.12 Individual Test Descriptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-18 14.3 Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-1 14.3.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-1 14.3.2 Tier 1 Design Description and Inspections, Tests, Analyses, and Acceptance Criteria First Principles . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-1 14.3.3 Organization of Tier 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-7 14.3.4 Tier 1 Chapter 1, Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-11 14.3.5 Tier 1 Chapter 2, Unit-Specific Structures, Systems, and Components Design Descriptions and Inspections, Tests, Analyses, and Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-11 14.3.6 Tier 1 Chapter 3, Shared Structures, Systems, and Components and Non-Structures, Systems, and Components Design Descriptions and Inspections, Tests, Analyses, and Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . 14.3-12 14.3.7 Tier 1 Chapter 4, Interface Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-12 14.3.8 Tier 1 Chapter 5, Site Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-12 Tier 2 i Revision 1

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 14.2-1: Spent Fuel Pool Cooling System Test # 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-19 Table 14.2-2: Pool Cleanup System Test # 2. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-21 Table 14.2-3: Reactor Pool Cooling System Test # 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-23 Table 14.2-4: Pool Surge Control System Test #4. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-24 Table 14.2-5: Ultimate Heat Sink Test # 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-26 Table 14.2-6: Pool Leak Detection System Test # 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-27 Table 14.2-7: Reactor Component Cooling Water System Test # 7 . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-28 Table 14.2-8: Chilled Water System Test # 8. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-30 Table 14.2-9: Auxiliary Boiler System Test # 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-32 Table 14.2-10: Circulating Water System Test # 10 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-34 Table 14.2-11: Site Cooling Water System Test # 11 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-36 Table 14.2-12: Potable Water System Test # 12. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-39 Table 14.2-13: Utility Water System Test # 13 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-40 Table 14.2-14: Demineralized Water System Test # 14 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-42 Table 14.2-15: Nitrogen Distribution System Test # 15 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-44 Table 14.2-16: Service Air System Test # 16 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-45 Table 14.2-17: Instrument Air System Test # 17 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-46 Table 14.2-18: Control Room Habitability System Test # 18 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-48 Table 14.2-19: Normal Control Room HVAC System Test # 19. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-51 Table 14.2-20: Reactor Building HVAC System Test # 20 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-55 Table 14.2-21: Radioactive Waste Building HVAC System Test # 21 . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-59 Table 14.2-22: Turbine Building Ventilation Test # 22 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-61 Table 14.2-23: Radioactive Waste Drain System Test # 23. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-63 Table 14.2-24: Balance-of-Plant Drains Test # 24 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-65 Table 14.2-25: Fire Protection System Test # 25 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-69 Table 14.2-26: Fire Detection Test # 26 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-71 Table 14.2-27: Main Steam Test # 27. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-72 Table 14.2-28: Feedwater System Test # 28 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-73 Table 14.2-29: Feedwater Treatment Test # 29 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-76 Table 14.2-30: Condensate Polisher Resin Regeneration System Test # 30 . . . . . . . . . . . . . . . . . . . 14.2-77 Table 14.2-31: Heater Vents and Drains Test # 31 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-79 Table 14.2-32: Condenser Air Removal System Test # 32 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-81 Tier 2 ii Revision 1

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 14.2-33: Turbine Generator Test # 33 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-83 Table 14.2-34: Turbine Lube Oil System Test # 34 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-86 Table 14.2-35: Liquid Radioactive Waste System Test # 35. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-88 Table 14.2-36: Gaseous Radioactive Waste System Test # 36. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-92 Table 14.2-37: Solid Radioactive Waste System Test # 37 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-94 Table 14.2-38: Chemical and Volume Control System Test # 38 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-97 Table 14.2-39: Boron Addition System Test # 39 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-101 Table 14.2-40: Module Heatup System Test # 40 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-103 Table 14.2-41: Containment Evacuation System Test # 41 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-104 Table 14.2-42: Containment Flooding and Drain System System Test # 42 . . . . . . . . . . . . . . . . . .14.2-107 Table 14.2-43: Containment System Test # 43. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-109 Table 14.2-44: Control Rod Drive System Flow-Induced Vibration Test # 44 . . . . . . . . . . . . . . . . .14.2-110 Table 14.2-45: Reactor Vessel Internals Flow-Induced Vibration Test # 45 . . . . . . . . . . . . . . . . . . .14.2-111 Table 14.2-46: Reactor Coolant System Test # 46 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-112 Table 14.2-47: Emergency Core Cooling System Test # 47 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-113 Table 14.2-48: Decay Heat Removal System Test # 48 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-114 Table 14.2-49: Incore Instrumentation Test # 49 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-115 Table 14.2-50: Module Assembly Equipment Test # 50 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-116 Table 14.2-51: Fuel Handling Equipment System Test # 51 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-117 Table 14.2-51a: FHE System Interlock Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-119 Table 14.2-52: Reactor Building Cranes Test # 52. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-120 Table 14.2-52a: RBC System Interlock Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-123 Table 14.2-53: Process Sampling System Test # 53 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-124 Table 14.2-54: 13.8kV and Switchyard System Test # 54 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-127 Table 14.2-55: Medium Voltage AC Electrical Distribution System Test # 55 . . . . . . . . . . . . . . . . .14.2-129 Table 14.2-56: Low Voltage AC Electrical Distribution System Test # 56 . . . . . . . . . . . . . . . . . . . . .14.2-131 Table 14.2-57: Highly Reliable DC Power System Test # 57. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-133 Table 14.2-58: Normal DC Power System Test # 58 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-135 Table 14.2-59: Backup Power Supply Test # 59 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-137 Table 14.2-60: Plant Lighting System Test # 60. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-139 Table 14.2-61: Module Control System Test # 61 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-141 Table 14.2-62: Plant Control System Test # 62. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-142 Tier 2 iii Revision 1

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 14.2-63: Module Protection System Test #63. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-143 Table 14.2-64: Plant Protection System Test # 64. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-151 Table 14.2-65: Neutron Monitoring System Test # 65. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-152 Table 14.2-66: Safety Display and Indication Test # 66. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-153 Table 14.2-67: Fixed Area Radiation Monitoring System Test # 67. . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-156 Table 14.2-68: Communication System Test # 68. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-157 Table 14.2-69: Seismic Monitoring System Test # 69 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-159 Table 14.2-70: Hot Functional Testing Test # 70 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-160 Table 14.2-71: Module Assembly Equipment Bolting Test # 71 . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-163 Table 14.2-72: Steam Generator Flow-Induced Vibration Test # 72. . . . . . . . . . . . . . . . . . . . . . . . . .14.2-164 Table 14.2-73: Security Access Control Test # 73 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-165 Table 14.2-74: Security Detection and Alarm-Test # 74 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-166 Table 14.2-75: Initial Fuel Loading Precritical Test (Test #75). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-167 Table 14.2-76: Initial Fuel Load Test (Test #76) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-168 Table 14.2-77: Reactor Coolant System Flow Measurement Test (Test #77). . . . . . . . . . . . . . . . . .14.2-169 Table 14.2-78: NuScale Power Module Temperatures Test (Test #78). . . . . . . . . . . . . . . . . . . . . . . .14.2-170 Table 14.2-79: Primary and Secondary System Chemistry Test (Test #79). . . . . . . . . . . . . . . . . . . .14.2-171 Table 14.2-80: Control Rod Drive System - Manual Operation, Rod Speed, and Rod Position Indication Test (Test #80) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-172 Table 14.2-81: Control Rod Assembly Drop Time Test (Test #81) . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-173 Table 14.2-82: Pressurizer Spray Bypass Flow Test (Test #82) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-174 Table 14.2-83: Initial Criticality Test (Test #83). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-175 Table 14.2-84: Post - Critical Reactivity Computer Checkout Test (Test #84) . . . . . . . . . . . . . . . . .14.2-176 Table 14.2-85: Low Power Test Sequence (Test #85). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-177 Table 14.2-86: Determination of Zero-Power Physics Testing Range Test (Test #86) . . . . . . . . .14.2-178 Table 14.2-87: All Rods Out Boron Endpoint Determination Test (Test #87) . . . . . . . . . . . . . . . . .14.2-179 Table 14.2-88: Isothermal Temperature Coefficient Measurement Test (Test #88) . . . . . . . . . . .14.2-180 Table 14.2-89: Bank Worth Measurement Test (Test #89) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-181 Table 14.2-90: Power-Ascension Test (Test #90). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-182 Table 14.2-91: Core Power Distribution Map Test (Test #91) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-183 Table 14.2-92: Neutron Monitoring System Power Range Flux Calibration Test (Test

92) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-184 Tier 2 iv Revision 1

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 14.2-93: Reactor Coolant System Temperature Instrument Calibration Test (Test

93) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-185 Table 14.2-94: Reactor Coolant System Flow Calibration Test (Test #94). . . . . . . . . . . . . . . . . . . . .14.2-186 Table 14.2-95: Radiation Shield Survey Test (Test #95). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-187 Table 14.2-96: Reactor Building Ventilation System Capability (Test #96). . . . . . . . . . . . . . . . . . . .14.2-188 Table 14.2-97: Thermal Expansion Test (Test #97) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-189 Table 14.2-98: Control Rod Assembly Misalignment (Test #98) . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-190 Table 14.2-99: Steam Generator Level Control Test (Test #99) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-191 Table 14.2-100: Ramp Change in Load Demand (Test #100) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-192 Table 14.2-101: Step Change in Load Demand Test (Test #101) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-193 Table 14.2-102: Loss of Feedwater Heater Test (Test #102). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-194 Table 14.2-103: 100 Percent Load Rejection Test (Test #103). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-195 Table 14.2-104: Reactor Trip from 100 Percent Power Test (Test #104) . . . . . . . . . . . . . . . . . . . . . . .14.2-196 Table 14.2-105: Island Mode Test for NuScale Power Module #1(Test #105) . . . . . . . . . . . . . . . . . .14.2-197 Table 14.2-106: Island Mode Test for Multiple NuScale Power Modules (Test #106) . . . . . . . . . . .14.2-198 Table 14.2-107: Remote Shutdown Workstation Test (Test #107) . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-199 Table 14.2-108: NuScale Power Module Vibration Test (Test #108) . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-200 Table 14.2-109: List of Test Abstracts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-201 Table 14.2-110: ITP Testing of New Design Features . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .14.2-204 Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-14 Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference . . . . . . . . 14.3-57 Tier 2 v Revision 1

NuScale Final Safety Analysis Report Verification Programs CHAPTER 14 INITIAL TEST PROGRAM AND INSPECTIONS, TESTS, ANALYSES, AND ACCEPTANCE CRITERIA 14.0 Verification Programs Verification programs include the initial test programs for the NuScale Power, LLC (NuScale)

Power Plant. The initial test programs are comprised of preoperational tests, initial fuel loading, initial criticality, low-power tests, and power-ascension tests. The verification programs ensure that the as-built facility configuration and operation comply with the approved plant design and applicable regulations.

The verification programs also include Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC). The methodology associated with developing ITAAC is described in Section 14.3. The ITAAC are presented in Tier 1.

The initial test program addresses structures, systems, and components and design features for both the nuclear portion of the facility and the balance-of-plant. The initial test program contains information that:

addresses the major phases of the test program including preoperational tests, initial fuel loading, initial criticality, low-power tests, and power-ascension tests, including scope and general plans for demonstrating that due consideration has been given to matters that normally require advance planning.

demonstrates that an adequate number of qualified personnel support the program.

demonstrates the adequacy of administrative controls to govern the conduct of the program.

allows plant staff the ability to train using the plants operating procedures.

demonstrates and verifies the adequacy of plant operating and emergency procedures to the extent practicable during the period of the initial test program.

allows for the verification of functional requirements.

demonstrates sequence of testing such that the safety of the plant does not depend on untested structures, systems, and components.

Tier 2 14.0-1 Revision 1

NuScale Final Safety Analysis Report Specific Information to be Addressed for the Initial Plant Test Program 14.1 Specific Information to be Addressed for the Initial Plant Test Program The initial test program establishes procedures and controls used to conduct and evaluate the results of tests as described in Section 14.2 and to satisfy the relevant requirements of the following regulations:

10 CFR 30.53, as it relates to testing radiation detection equipment and monitoring instruments

10 CFR 50.34(b)(6)(iii), as it relates to providing information associated with preoperational testing and initial operations

Section XI of Appendix B to 10 CFR Part 50, as it relates to test programs to demonstrate that systems, structures, and components will perform satisfactorily

Option A or Option B of Appendix J of 10 CFR Part 50, as it relates to preoperational leakage rate testing

10 CFR 52.79 as it pertains to preoperational testing and initial operations

Subpart A, Subpart B, and Subpart C of 10 CFR Part 52 as they relate to the Inspections, Tests, Analyses, and Acceptance Criteria that the applicant must submit Tier 2 14.1-1 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program 14.2 Initial Plant Test Program 14.2.1 Summary of Initial Test Program and Objectives The Initial Test Program (ITP) consists of a series of preoperational and startup tests.

Preoperational testing is conducted following completion of construction testing but prior to fuel load. Completion of preoperational testing is necessary to ensure the overall plant is ready for fuel loading and startup testing of a NuScale Power Module (NPM).

Startup tests of an NPM are performed following the completion of preoperational testing.

Startup testing includes the following:

initial fuel loading and pre-critical testing

initial criticality testing

low-power testing

power-ascension testing Startup testing is performed to confirm the design bases of the NPM and to demonstrate, to the extent practical, that the NPM will operate in accordance with its design and is capable of responding to anticipated transients and postulated accidents as described in Section 15.0.

The objectives of the ITP are to

provide assurance that structures, systems, and components (SSC) operate in accordance with their design.

provide assurance that construction and installation of equipment in the facility has been completed in accordance with the design. Verification of design requirements is also performed as part of construction testing phase of the ITP.

demonstrate to the extent practical the validity of analytical models used to predict plant responses to anticipated transients and postulated accidents, and to demonstrate to the extent practical the correctness and conservatism of assumptions used in those models.

familiarize the plant's operating and technical staff with the operation of the facility.

perform testing to the extent practical using the plant conditions that simulate the actual operating, abnormal operating occurrences, and emergency conditions to which the SSC may be subjected.

verify to the extent practical by trial use that the facility operating procedures, surveillance procedures and emergency procedures are adequate.

verify that interfaces and system and component interactions are in accordance with the design.

complete and document the ITP testing required to satisfy preoperational and startup testing requirements and Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) testing requirements.

Tier 2 14.2-1 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Preoperational and startup testing is performed on those SSC that are:

relied upon for safe shutdown and cooldown of the NPM under normal conditions and for maintaining a safe condition for an extended shutdown period.

relied upon for safe shutdown and cooldown of the NPM under transient and postulated accident conditions and for maintaining a safe condition for an extended shutdown period following such conditions.

relied upon for establishing conformance with safety limits or limiting conditions for operation that are included in the technical specifications (TS).

assumed to function or for which credit is taken in the accident analysis as described in Chapter 15.

used to process, store, control, or limit the release of radioactive materials.

relied upon to maintain their structural integrity during normal operation, anticipated transients, simulated test parameters, and design basis event conditions to avoid damage to safety-related SSC.

identified as risk-significant in the probabilistic risk assessment.

The ITP is implemented consistent with the requirements of Section XI of 10 CFR 50 Appendix B. Implementation of the ITP ensures that the testing required to demonstrate that SSC perform satisfactorily in service, are identified and performed in accordance with written test procedures that incorporate the requirements and acceptance limits in the applicable design documents.

Leakage rate testing of the NPM and related systems and components penetrating the containment pressure boundary is described in Section 6.2. Leakage rate testing test abstracts are presented in Section 14.2.12.

The methodology associated with the development of the ITAAC necessary to demonstrate that the facility has been constructed and will be operated in conformity with the final safety analysis report and the applicable Nuclear Regulatory Commission (NRC) regulations is presented in Section 14.3.

NuScale Power Plant compliance with the proposed technical resolution of unresolved safety issues and medium- and high-priority generic safety issues that are identified in NUREG-0933 are addressed in Section 1.9.3. Operating experience insights are addressed in Section 1.9.4 and Section 14.2.8. Compliance with technically relevant portions of the Three Mile Island requirements are addressed in Section 1.9.5.

14.2.1.1 Construction and Installation The objective of construction and installation tests are to verify that on a system basis that the system is constructed and installed in accordance with design requirements.

Construction tests include but are not limited to:

initial instrument calibration

flushing

cleaning Tier 2 14.2-2 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program

hydrostatic pressure tests

wiring

continuity and separation checks

functional tests of components 14.2.1.2 Preoperational Test Phase Objectives Preoperational tests are performed to demonstrate that SSC operate in accordance with design requirements so that initial fuel loading, initial criticality, and subsequent power operation can be safely undertaken. The objectives of the preoperational test phase are to

demonstrate that SSC will perform their functions in accordance with their design during the preoperational test phase.

verify and demonstrate expected operation following a loss of power sources and in degraded modes for which the systems are designed to remain operational.

test the backup power supply system (BPSS) to ensure that backup sources of alternating current (AC) electrical power are available when the normal AC power sources are not available.

verify and demonstrate the operational readiness of valves and dynamic restraints before relying on those components to perform their safety functions.

perform inspections or testing for flow-induced vibration loads on components that must maintain their structural integrity.

obtain baseline test and operating data on equipment and systems for future reference.

operate equipment for a sufficient period of time to achieve normal equilibrium conditions (e.g., temperatures and pressures) so that design, manufacturing, and installation defects can be detected and corrected.

ensure to the extent practical plant systems operate properly on an integrated basis.

evaluate normal, abnormal, and emergency operating procedures to the extent practical.

demonstrate equipment performance.

test, as appropriate, manual operation and automatic operation of systems and their components.

test the proper functioning of controls, permissives, interlocks, and equipment protective devices for which malfunction or premature actuation may shut down or defeat the operation of systems or equipment.

provide the plant operating staff with the opportunity to obtain practical experience in the operation and maintenance of equipment and systems including instrument calibrations and functional tests of components.

demonstrate equipment performance is satisfactory to proceed to initial fuel loading and initial criticality.

Tier 2 14.2-3 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Test abstracts associated with preoperational testing are included in Section 14.2.12.

14.2.1.3 Startup Test Phase Objectives 14.2.1.3.1 Initial Fuel Loading and Pre-Critical Tests This phase of testing is performed in order to ensure that initial fuel loading of an NPM can be accomplished in an orderly and safe manner. A description of the fuel loading process is presented Section 14.2.10.2. The objectives of the initial fuel loading and pre-critical tests are to:

conduct initial fuel loading cautiously to preclude inadvertent criticality.

Establish and follow specific safety measures, such as:

ensuring that the applicable TS requirements and other prerequisites have been satisfied continuous monitoring of the neutron flux throughout core loading so that changes in the multiplication factor are observed verifying that the fuel and control components have been properly installed

establish that the required SDM exists, without achieving criticality

establish the functionality of plant systems and components, including reactivity control systems and other systems and components necessary to ensure the safety of plant personnel and the public in the event of errors or malfunctions

confirm the proper operation of plant systems and design features that could not be completely tested during preoperational testing

confirm interdependent effects among the safety features of the design are acceptable 14.2.1.3.2 Initial Criticality The objectives associated with the initial criticality phase of the startup testing program are to achieve initial criticality in a safe and controlled manner. In order to meet this objective the following are performed:

The initial approach to criticality is performed in a deliberate and orderly manner using the same rod withdrawal sequences and patterns that will be used during subsequent startups.

The neutron flux levels are continuously monitored and periodically evaluated.

A neutron count rate of at least 1/2 counts per second should register on the startup channels before startup begins, and the signal to noise ratio should be known to be greater than 2.

The systems required for startup or protection of the plant, including the reactor protection system and engineered safety features (ESFs), are operable and in a state of readiness.

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NuScale Final Safety Analysis Report Initial Plant Test Program

The control rod or poison removal sequence is accomplished using approved plant procedures.

The reactor achieves initial criticality by boron dilution. Control rods are withdrawn before dilution begins.

The control rod insertion limits defined in the technical specifications are observed and followed.

Criticality predictions for boron concentration and control rod positions are provided.

The reactivity addition sequence is prescribed, and plant procedures require a cautious approach to achieving criticality to prevent passing through criticality in a period shorter than approximately 30 seconds (<1 decade per minute).

A description of the process followed to achieve initial criticality is provided in Section 14.2.10.3.

14.2.1.3.3 Low - Power Testing Following criticality, low-power testing is performed. The objectives associated with performing low-power testing are to

confirm the design and validate analytical models.

verify the correctness of assumptions used in the safety analyses.

confirm the functionality of plant systems and design features that could not be completely tested during the preoperational test phase because of the lack of an adequate heat source for the reactor coolant and main steam systems.

14.2.1.3.4 Power-Ascension Testing Following low-power testing, power-ascension testing is performed. Power-ascension testing is performed to bring the reactor to full power with testing at power levels of approximately 25 percent, 50 percent, 75 percent, and 100 percent.

The objectives associated with performing power-ascension testing are to

achieve reactor full power in a safe and controlled manner.

demonstrate that the plant operates in accordance with its design bases during normal steady-state conditions and, to the extent practical, during and following anticipated transients.

validate models used to predict plant response.

demonstrate the ability of major or principal plant control systems to automatically control process variables within design limits.

demonstrate that the facility's integrated dynamic response is in accordance with design for plant events such as reactor scram, turbine trip, and loss of feedwater heaters or pumps.

Tier 2 14.2-5 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program 14.2.2 Organization and Staffing COL Item 14.2-1: A COL applicant that references the NuScale Power Plant design certification will describe the site-specific organizations that manage, supervise, or execute the Initial Test Program, including the associated training requirements.

14.2.3 Test Procedures 14.2.3.1 Initial Test Program Procedures Test procedures are developed and reviewed by individuals with the appropriate technical background and expertise. Once the test procedures have been developed they are reviewed by plant management personnel who upon acceptance designate the procedures as final.

Input from the principal design organization is utilized to establish the test objectives and acceptance criteria for the system. Operating experience, as discussed in Section 14.2.8 is used in the development of test procedures.

Test procedure testing and acceptance criteria are founded upon the information contained in design specifications, design documents, the Final Safety Analysis Report, and regulatory documents. A test procedure is prepared for each specific system test to be performed during the test program.

Preoperational and startup testing procedures include checklists and signature blocks to control the sequence and performance of testing. The administrative controls associated with test procedure development address the following:

test procedure format

application, to the extent practical, of normal plant operating procedures, emergency operating procedures, and surveillance procedures in support of test procedure development

test procedure review and approval

test procedure change and revision The content of each test procedure addresses

objectives.

detailed step-by-step procedures specifying how testing is to be performed.

special precautions.

test instrumentation.

test equipment calibration.

initial test conditions, including provisions to perform testing under environmental conditions as close as practical to those the equipment will experience in both normal and accident situations.

methods to direct and control test performance.

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NuScale Final Safety Analysis Report Initial Plant Test Program

acceptance criteria by which testing is evaluated. Acceptance criteria account for measurement errors and uncertainties associated with normal operation as well as operation during transients and accidents. Acceptance criteria are biased conservatively. In some cases the acceptance criteria is qualitative. Where applicable, quantitative values, with appropriate tolerances, are used as acceptance criteria.

test prerequisites including as necessary prerequisite statements to ensure that nonstandard arrangements are restored to their normal status after the test is completed (for example, electric jumper cable use does not invalidate electrical separation; jumper cables are removed following testing; valve configurations, and instrument settings are returned to their normal orientations and settings).

identification of the data to be collected and the method of documentation.

actions to take if unanticipated errors or malfunctions occur while testing.

remedial actions to take if acceptance criteria are not satisfied.

actions to take if an unexpected or unanalyzed condition occurs.

14.2.3.2 Graded Approach to Testing The ITP allows for the application of a graded approach to testing. The graded approach to testing is founded in the requirements of General Design Criterion 1, Quality Standards and Records, of Appendix A to 10 CFR Part 50 that requires, in part, that SSC important to safety shall be tested to quality standards commensurate with the importance of the safety functions to be performed. Criterion XI of Appendix B to 10 CFR Part 50 also includes a graded approach for important to safety SSC in the Quality Assurance Program. The administrative requirements that govern the conduct of the test program (e.g., test program objectives, organizational elements, personnel qualifications, evaluation and approval of test results, and test records retention) contain provisions that allow for testing of SSC in a manner commensurate with the safety significance of the SSC within its scope. This provides a systematic approach to the defense-in-depth concept. This concept dictates that the plant be designed, constructed, and tested to (1) provide for safe normal operation, (2) ensure that, in the event of errors, malfunctions, and off-normal conditions, the reactor protection systems and other design features will mitigate the event or limit its consequences to defined and acceptable levels, and (3) ensure that adequate safety margin exists for events of extremely low probability or arbitrarily postulated hypothetical events without substantial reduction in the safety margin for the protection of public health and safety.

Application of the graded approach to testing provides reasonable assurance that the SSC being tested will perform satisfactorily while accomplishing the testing in a cost-effective manner. The administrative requirements that govern the conduct of the test program allow for the preparation of documentation (such as procedures and records) associated with testing to be prepared commensurate with the importance to safety of the SSC being tested.

During the SSC classification process, the subject matter expert identified all functions of the system. Each of these functions was compared to safety functional requirements Tier 2 14.2-7 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program and regulatory functional requirements to establish a functional hierarchy. This hierarchy established a supporting to supported relationship between the systems and tied it to a set of plant functions as described in Section 17.4 to identify a classification for the functions. The functions were categorized as A1 (safety-related, risk-significant),

A2 (safety-related, not risk-significant), B1 (nonsafety-related, risk-significant), or B2 (nonsafety-related, not risk-significant). This safety significance evaluation was the basis for the graded approach in the ITP.

The hierarchy in the NuScale approach to preoperational testing is:

Testing of active, safety-related system functions (A1 or A2 functions)

Testing of active, non-safety-related functions which require ITAAC verification (B1 and B2)

Testing of active non-safety-related functions which do not require ITAAC verification (B1 and B2)

The preoperational test abstracts contained in Table 14.2-1 through Table 14.2-69 provide a definition of the test scope for each system by listing the associated active system functions and their safety categorization. The test abstract also provides system functions tested by another test abstract, thereby providing an "inventory" of all testable system functions.

Table 17.4-1 contains a list of all A1 and B1 system functions. All active, safety-related A1 functions are tested by the safety-related module protection system (MPS) logic testing found in Table 14.2-63. The remaining safety-related functions categorized as A2 are also tested by the MPS test abstract. The NuScale graded approach provides for testing of A2 functions to the same rigor as A1 functions.

As indicated by Table 14.2-63, all active, safety-related functions are one of the following types:

provides safety-related instrument information signals to MPS

removes electrical power to the control rod drive

removes electrical power to the pressurizer heaters

removes electrical power to the trip solenoids of safety-related valves

closes safety-related valves The MPS test abstract also describes testing of the following safety-related design features:

MPS response to loss of electrical power

MPS operating bypasses and permissives

Safety-related containment isolation valve response time

MPS safety-related sensor response time Table 14.3-1 and Table 14.3-2 identify all ITAAC by its unique ITAAC number. The tables provide a discussion of the ITAAC, including a reference to a verifying preoperational Tier 2 14.2-8 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program test abstract, if required. This results in a cross-reference between Tier 1 ITAAC and the associated Tier 2 test abstract. Table 14.3-1 or Table 14.3-2 is annotated with the unique ITAAC number entered in the acceptance criteria column of the associated test abstract.

Table 14.2-63 identifies that the acceptance criteria of the all MPS test abstracts which test a safety-related function have an associated ITAAC.

Preoperational testing of non-safety-related systems is necessary to verify ITAAC for the following design features. The acceptance criteria of the associated test abstract acceptance criteria are annotated with the ITAAC number.

Radiation isolation

Battery room ventilation for hydrogen control

Control room building and reactor building differential pressure

Control room envelope design features

Post-accident monitoring (PAM) signals

Turbine trip

Fire protection pump flow

Plant lighting illumination in the main control room, remote shutdown, and for post-fire shutdown

Important Human Actions for CFD addition of water to containment

Important Human Actions for CVC addition of water to the reactor coolant system Credit is taken for the logic testing performed for the nonsafety-related module control system (MCS) described in Section 7.0.4.5, and the nonsafety-related plant control system (PCS) described in Section 7.0.4.6. Therefore, if the component is controlled by MCS or PCS, the component-level logic testing in the preoperational test is limited to the testing of component-level design features described below (if the design feature is applicable to the system) unless the preoperational test verifies an ITAAC. The component tests are standardized to provide the same level of test detail across all systems. This graded approach does not affect system-level tests which require integrated system operation. The standardized component tests are:

Remote operation of equipment.

Manual control of variable-speed pump or fan.

Automatic start of standby pump or fan.

Automatic operation of pump recirculation valve.

Pump start does not create a water hammer.

Remote operation of valve or damper.

Valve or damper fails to its safe position on loss of air.

Valve or damper fails to its safe position on loss of electrical power to its solenoid.

Damper or fan responds to fire or smoke alarm.

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NuScale Final Safety Analysis Report Initial Plant Test Program

Equipment response to automatic signals to protect plant equipment.

Automatic operation of tank or basin level control valve.

Local grab sample can be obtained from a system grab sample device.

Automatic bus transfer via bus tie breaker.

System instrument calibration.

Each instrument is monitored in the MCR and the remote shutdown station (RSS), if the signal is designed to be displayed in the RSS. (Test not required if the instrument calibration verified the MCR and RSS display.)

Equipment protection logic 14.2.3.3 Testing of First-of-a-Kind Design Features First-of-a-kind (FOAK) tests are new, unique, or special tests used to verify design features that are being reviewed for the first time by the NRC. The NuScale Power Plant contains design features which are new and unique and have not been tested previously; therefore, testing of these design features is treated as FOAK. For the FOAK tests, the testing frequency is specified in the test abstract. The NuScale comprehensive vibration assessment program is a FOAK program. The program is implemented consistent with the requirements of the NuScale "Comprehensive Vibration Assessment Program (CVAP) Technical Report", TR-0716-50439. The CVAP is addressed in Section 3.9.2.

The following ITP test abstracts describe the on-site CVAP testing of FOAK design features:

Table 14.2-44: Control Rod Drive System Flow-Induced Vibration Test #44

Table 14.2-45: Reactor Vessel Internals Flow-Induced Vibration Test #45

Table 14.2-75: Steam Generator Flow-Induced Vibration Test #72.

Table 14.2-108: NuScale Power Module Vibration Test #108 The test results for the CVAP program testing of the first NPM are to inform the required CVAP testing on subsequent NPMs as described in Section 6.0 of TR-0716-50439. All other ITP testing of FOAK design features is performed for each NPM.

Table 14.2-110 provides a summary of the ITP testing (i.e., preoperational and startup testing) for new design features. Each test will be performed for all NPMs.

Section 1.5.1 contains a description of testing programs which have been completed or are currently in progress for NuScale design features for which applicable data or operational experience did not previously exist. The section describes tests specific to fuel design, steam generator (SG) and control rod assemblies.

14.2.4 Conduct of the Test Program The ITP activities are controlled by administrative procedures contained within the Startup Administrative Manual.

Tier 2 14.2-10 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program COL Item 14.2-2: A COL applicant that references the NuScale Power Plant design certification is responsible for the development of the Startup Administration Manual that will contain the administrative procedures and requirements that control the activities associated with the Initial Test Program. The COL applicant will provide a milestone for completing the Startup Administrative Manual and making it available for NRC inspection.

Administrative controls are established to ensure that the designated construction-related inspections and tests are completed prior to initiating preoperational testing. In addition controls are established to ensure completion of preoperational testing prior to initiating startup testing. Administrative controls address adherence to approved test procedures during the conduct of the test program and the methods for effecting changes to approved test procedures.

The controls used to ensure that test prerequisites associated with each major phase of testing, as well as individual system or component testing are met, include requirements for performing inspections and checks, identification of test personnel, completing data forms or check sheets, and identification of dates of completion.

The controls provided to implement plant modification and repairs ensure that the required modifications and repairs are made. Retesting is conducted following modifications or repairs. Reviews of proposed facility modifications by designated design organizations is conducted prior to performing the modification or repair.

Controls are established to ensure that retesting that is required for modifications or maintenance remains in compliance with ITAAC commitments.

The documentation associated with the conduct of the test plan is captured and auditable.

14.2.5 Review, Evaluation, and Approval of Test Results Administrative procedures control the review and approval of preoperational and startup test results for each phase of the test program. This includes approval of test data for each major test phase before proceeding to the next test phase as well as approval of test data at each power test plateau (during the power-ascension phase) before increasing the power level. Test exceptions or results that do not meet acceptance criteria are identified to the responsible design organization as well as plant operations and plant technical staff and corrective actions and retests, as required, are performed.

These administrative procedures address the following:

notification of responsible design organizations when test acceptance criteria are not met

methods and schedules for approval of test data for each major phase

methods used for initial review of individual parts of multiple tests

technical evaluation of test results by qualified personnel and approval of such results by personnel in designated management positions Tier 2 14.2-11 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program

provisions to allow design organizations to participate in the resolution of design-related problems that result in, or contribute to, a failure to meet test acceptance criteria

provisions to retain test reports, including test procedures and results, as part of the plant historical records 14.2.6 Test Records Initial test program reports, test procedures and results are retained as part of the plant's historical record in accordance with 10 CFR 50.36, "Technical Specification," 10 CFR 50.71, "Maintenance of Records, Making of Reports," and 10 CFR 50, Appendix B, Criterion XVII, "Quality Assurance Records." The test reports include test results associated with the testing of SSC identified in the ITP. A summary of the startup testing is included in a startup report. This summary includes the following information:

description of the method and objectives for each test

comparison of applicable test data with the related acceptance criteria, including the systems' responses to major plant transients (such as reactor scram and turbine trip)

design and construction related deficiencies discovered during testing, and system modifications, the corrective actions required to correct those deficiencies, and the schedule for implementing the identified modifications and corrective actions

justification for acceptance of systems or components that are not in conformance with design predictions or performance requirements

conclusions about system or component adequacy

identity of test observers and recorders

type of observation

identifying numbers of test or measuring equipment

results of tests 14.2.7 Test Programs Conformance with Regulatory Guides The ITP conforms to Regulatory Guide (RG) 1.68, Revision 4 except for aspects that address specific SSC design features not in the design.

The following list of regulatory guides provides information used to supplement the information, recommendations, and guidance presented in RG 1.68 Rev 4 relative to testing of SSC:

RG 1.20 - Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing, Rev. 3

RG 1.29 - Seismic Design Classification for Nuclear Power Plants, Rev. 5

RG 1.45 - Guidance on Monitoring and Responding to Reactor Coolant System Leakage, Rev. 1

RG 1.68.1 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors, Rev. 2 Tier 2 14.2-12 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program

RG 1.68.2 - Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water- Cooled Nuclear Power Plants, Rev 2

RG 1.68.3 - Preoperational Testing of Instrument and Control Air Systems, Rev.1

RG 1.69 - Concrete Radiation Shields and Generic Shield Testing for Nuclear Power Plants, Rev. 1

RG 1.79 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors, Rev. 2

RG 1.118 - Periodic Testing of Electric Power and Protection Systems, Rev. 3

RG 1.128 - Installation Design and Installation of Vented Lead-Acid Storage Batteries for Nuclear Power Plants Rev. 2

RG 1.140 - Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Normal Atmosphere Cleanup Systems in Light-Water- Cooled Nuclear Power Plants, Rev. 2

RG 1.155 - Station Blackout, Rev.1

RG 8.38 - Control of Access to High and Very High Radiation Areas of Nuclear Power Plants, Rev. 1 14.2.8 Utilization of Reactor Operating and Testing Experience in Test Program Development The operational experience gained from pressurized-water and other reactor designs is factored into the design and testing.

Operations and technical staff review the following documents for information that can be included in the ITP:

NRC licensee event reports

NRC generic communications (i.e., inspection and enforcement bulletins, circulars, generic letters, administrative letters, information notices, and regulatory issue summaries)

Institute of Nuclear Power Operations issuances The administrative procedures control the review of reactor operating experience and its incorporation in the ITP.

14.2.9 Trial Use of Plant Operating Procedures, Emergency Procedures, and Surveillance Procedures Plant emergency, operating, and surveillance test procedures are, to the extent practical, developed, trial tested, and corrected during the ITP before fuel load to establish their adequacy. Trial testing of procedures is accomplished by having plant operators trained to these procedures to the extent practicable during the ITP. Following completion of trial testing these procedures are used as part of the ITP.

Tier 2 14.2-13 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Additionally, emergency, operating, and surveillance test procedures are incorporated into the plant reference simulator, which meets the requirements of 10 CFR 55.46(c), and are trial tested as part of the operator training program.

The administrative procedures control the trial use of approved plant operating procedures, emergency operating procedures, and surveillance procedures.

14.2.10 Initial Fuel Loading, and Initial Criticality Approved startup tests are used to control startup testing for initial fuel loading, pre-critical tests, initial criticality, low-power tests, and power-ascension tests in a controlled, deliberate, and safe manner. Technical specification compliance is met prior to initiation of startup testing. Startup test procedures are prepared based upon test abstracts provided in Section 14.2.12.

Startup tests procedures contain general provisions, precautions, prerequisites, and measures consistent with the requirements of RG 1.68 Rev. 4.

14.2.10.1 Initial Fuel Loading and Pre-Criticality Testing As part of the startup test program, initial fuel loading and pre-criticality testing are performed by first implementing the prerequisite and precautionary measures that are contained in test procedures and identified below:

technical specification compliance is met

successful completion of all ITAAC

actions to be taken in the event of unanticipated errors or malfunctions are clearly identified

completion of a review of preoperational test results (the Startup Administrative Manual contains administrative procedures to control the verification process for successful completion of preoperational tests required for fuel load)

review and status of design changes

review of retests that were performed due to preoperational test deficiencies

review of test exceptions 14.2.10.2 Initial Fuel Loading Initial fuel loading is conducted to preclude inadvertent criticality. Specific safety measures are followed including (1) ensuring that the applicable TS are met, (2) performing continuous monitoring of the neutron flux throughout core loading so that changes in the multiplication factor are observed, (3) establishing requirements for periodic data taking, and (4) independently verifying that the fuel and control components have been properly installed.

Predictions of core reactivity are prepared in advance of the initial fuel loading to aid in evaluating the measured responses to specified loading increments. Comparative data on neutron detector responses from previous loadings of essentially identical core Tier 2 14.2-14 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program designs may be used in lieu of these predictions. Criteria and requirements for actions to be taken if the measured results deviate from expected values are established prior to the initial fuel loading. In addition, prior to initial fuel loading the required shutdown margin (SDM) is confirmed.

To provide further assurance of safe loading, requirements for the functionality of plant systems and components are established, including reactivity control systems and other systems and components necessary to ensure the safety of plant personnel and the public in the event of errors or malfunctions. The initial core loading is directly supervised by a senior licensed operator having no other concurrent duties, and the loading operation is conducted in strict accordance with detailed approved procedures.

14.2.10.3 Initial Criticality Testing Control rods are withdrawn in the normal sequence to a configuration that does not violate the zero power rod insertion limits. Initial criticality is then achieved in a deliberate, orderly, and controlled fashion using boron dilution. Core neutron flux is continuously monitored during the approach to critical. Changes in reactivity are continuously monitored, and inverse multiplication plots are maintained and interpreted.

The following conditions exist prior to initial criticality:

A minimum crew is required to support initial criticality, including a senior reactor operator with no other concurrent duties who is in charge of the operation.

Critical rod position and boron concentration predictions are identified so that anomalies can be noted and evaluated.

Systems needed for startup are aligned and in proper operation.

Emergency systems are operable and in readiness.

TS compliance is met.

Nuclear instruments are calibrated.

Neutron count rate of at least 1/2 counts per second registers on startup channels before the startup begins.

Signal to noise ratio is greater than two.

Conservative startup rate limit (greater than approximately a 30-second period) is established.

High flux scram trips are set at their lowest value.

Implementation of the radiation monitoring program as it pertains to operation of radiation barriers, airborne radiation monitors, air sampling, as well as performance of baseline surveys before pulling control rods for the approach to critical.

Tier 2 14.2-15 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program 14.2.10.4 Low-Power Testing Following initial criticality, low-power tests (at less than 5 percent power) are conducted to (1) confirm the design and, to the extent practical, validate the analytical models, and verify the correctness or conservatism of assumptions used in the safety analyses for the facility, and (2) confirm the functionality of plant systems and design features that could not be completely tested during the preoperational test phase because of the lack of an adequate heat source for the reactor coolant and main steam systems.

Low-power testing is performed in a controlled manner in accordance with written procedures. The minimum crew required to support low-power testing is available in addition to a senior reactor operator with no other concurrent duties who is in charge of low-power testing operations. Low-power testing procedures include instructions and precautions necessary for conducting tests such as adherence to TS requirements, testing sequence, measurement to be taken and test conditions as well as actions to be taken in the event of unanticipated errors or malfunctions. These procedures provide direction for restoration to normal following the test.

Refer to Section 14.2.12 for a list of low-power tests.

COL Item 14.2-3: A COL applicant that references the NuScale Power Plant design certification will identify the specific operator training to be conducted during low-power testing related to the resolution of TMI Action Plan Item I.G.1, as described in NUREG-0660, NUREG-0694, and NUREG-0737.

14.2.10.5 Power-Ascension Tests Power-ascension testing is performed following the successful completion of low-power testing. Power-ascension testing is performed to bring the reactor to full power and while doing so performing major testing at power levels of approximately 25 percent, 50 percent, 75 percent, and 100 percent. The purpose of the testing is to demonstrate that the plant operates in accordance with its design bases during normal steady state conditions and, to the extent practicable, during and following anticipated transients as well as to demonstrate the validity of analytical models by comparing measured responses with predicted responses. Predicted responses are developed using real or expected values of attributes such as beginning of life core reactivity coefficients, flow rates, pressures, temperatures, and response times of equipment, as well as the actual status of the plant (not those values or plant conditions assumed for conservative evaluations of postulated accidents).

Tests and acceptance criteria are prescribed to demonstrate the ability of principal plant control systems to automatically control process variables within design limits.

Such tests are expected to provide assurance that the facility's integrated dynamic response is in accordance with the design for plant events such as reactor scram, turbine trip, and loss of feedwater heaters or pumps. The testing performed is sufficiently comprehensive to establish that the facility can operate in the operating modes for which it has been designed. Testing is not conducted in operating modes or plant configurations that have not been analyzed or that fall outside the range of Tier 2 14.2-16 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program assumptions used in analyzing postulated accidents described in the Final Safety Analysis Report.

Power-ascension testing is performed in a controlled manner in accordance with written procedures. The minimum crew required to support power-ascension testing is available in addition to a senior reactor operator with no other concurrent duties who is in charge of power-ascension testing operations. Power-ascension testing procedures include instructions and precautions necessary for conducting tests such as adherence to TS requirements, testing sequence, measurement to be taken and test conditions as well as actions to be taken in the event of unanticipated errors or malfunctions. These procedures provide direction for restoration to normal following the test.

Refer to Section 14.2.12 for a list of power-ascension tests.

The completed power-ascension testing program is reviewed at each plateau. Test results are evaluated and the required approvals are received before ascending to the next power level or test condition.

14.2.11 Test Program Schedule and Sequence Testing schedules are developed taking into account development and approval of plant procedures for use as part of the ITP.

Testing schedules are developed so that SSC that are required to prevent or mitigate the consequences of postulated accidents are tested prior to fuel loading.

Approved test procedures are submitted to the NRC approximately 60 days before their intended use or at least 60 days prior to fuel loading for fuel loading and startup test procedures. The NRC is notified of test procedure changes prior to performance.

Test procedures are essentially identical for each NPM. SSC identification numbering is specific to each NPM.

For individual startup tests, test requirements are completed in accordance with plant TS requirements associated with SSC functionality before changing plant modes.

Testing required to be completed prior to fuel load that is intended to satisfy the requirements for completing ITAAC is identified and documented as such.

Vibration testing that is performed at the factory is performed in accordance to the requirements of the NuScale "Comprehensive Vibration Assessment Program" as described in the "Comprehensive Vibration Assessment Program (CVAP) Technical Report," TR-0716-50439. The technical report contains a schedule for the CVAP testing. Test results are verified following power-ascension testing. See Section 3.9.2 for information pertaining to the CVAP.

The sequential schedule for individual startup tests establishes, insofar as practicable, that test requirements are completed prior to exceeding 25 percent power for the plant SSC that are relied upon to prevent, limit, or mitigate the consequences of postulated Tier 2 14.2-17 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program accidents. The schedule establishes that, insofar as practicable, the sequencing of testing is accomplished as early in the test program as feasible and that the safety of the plant is not dependent on the performance of untested systems, components, or features. Startup test data is reviewed and approved prior to moving onto the next power plateau. Startup testing is discussed in Section 14.2.1.3.

The NuScale Power Plant is comprised of up to 12 NPMs. A schedule is developed for startup of each NPM. Preoperational and startup testing schedule considerations include:

preoperational test schedule duration will be greatest for the first NPM because the first NPM will require testing of systems common to other NPMs

preoperational and startup test schedule duration should decrease for each successive NPM due to increase in personnel experience and refinement of test procedures

scheduling such that overlapping test program schedules will not result in significant divisions of responsibilities or dilute staff provided to implement the test program

plant safety will not be dependent on the performance of untested SSC during the startup test program Refer to Section 21.3.3 for information pertaining to phased construction and testing activities due to addition of individual NPMs.

COL Item 14.2-4: A COL applicant that references the NuScale Power Plant design certification will provide a schedule for the Initial Test Program.

14.2.12 Individual Test Descriptions Individual test abstracts are provided in Table 14.2-1 through Table 14.2-108. Table 14.2-109 provides a listing of the test abstracts. Each abstract identifies each test by title, identifies test objectives, prerequisites, test methods, and acceptance criteria. Detailed preoperational and startup test procedures are developed using these test abstracts.

The test abstracts identify pertinent precautions for individual tests, as necessary (e.g.,

minimum flow requirements or reactor power level that must be maintained).

Tier 2 14.2-18 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-1: Spent Fuel Pool Cooling System Test # 1 Preoperational test is required to be performed once.

The SFPCS is described in Section 9.1.3.2.1. SFPCS functions are not verified by SFPCS tests. SFPCS functions verified by another test is:

System Function System Function Categorization Function Verified by Test #

The spent fuel pool cooling system nonsafety-related Test #2-1 (SFPCS) supports the pool cleanup system (PCUS) by providing fuel pool water for purification of the ultimate heat sink (UHS).

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each SFPCS remotely-operated Operate each valve from the main MCR display and local, visual valve can be operated remotely. control room (MCR) and local control observation indicate each valve fully panel (if design has local valve control) opens and fully closes.

ii. Verify each SFPCS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each SFPCS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each SFPCS pump can be Align the SFPCS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

v. Verify each SFPCS pump Align the SFPCS to allow for pump MCR display and local, visual automatically stops to protect the operation. Place a pump in service. observation indicate each pump stops.

pump. Initiate a simulated stop signal for the following system conditions.

i. Low pump suction pressure ii. High pump discharge pressure.

vi. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a SFPCS grab sample through the grab sampling device. obtained.

device.

vii. Verify each SFPCS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the instrument signal from each SFPCS on an MCR workstation or recorded remote shutdown station (RSS), if the transmitter. by the applicable control system signal is designed to be displayed in historian.

the RSS. ii. The instrument signal is displayed (Test not required if the instrument on an RSS workstation or recorded calibration verified the MCR and RSS by the applicable control system display) historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-19 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-1: Spent Fuel Pool Cooling System Test # 1 (Continued)

System Level Tests None Tier 2 14.2-20 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-2: Pool Cleanup System Test # 2 Preoperational test is required to be performed once.

The PCUS is described in Section 9.1.3.2.3 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The SFPCS supports the PCUS by nonsafety-related Test #2-1 providing spent fuel pool water for purification of the UHS.

2. The reactor pool cooling system nonsafety-related Test #2-1 (RPCS) supports the PCUS by providing reactor pool water for purification of the UHS.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify a pump curve test has been completed and approved for the RPCS pumps.

iii. Verify a pump curve test has been completed and approved for the SFPCS pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each PCUS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

ii. Verify each PCUS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each PCUS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a PCUS grab sample through the grab sampling device. obtained.

device.

v. Verify each PCUS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each PCUS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-21 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-2: Pool Cleanup System Test # 2 (Continued)

System Level Test #2-1 Test Objective Test Method Acceptance Criteria Verify the PCUS demineralizers are i. Place the SFPCS in service to flow i. a. The MCR indication for SFPCS protected against high water through a pool cleanup filter and a pump flow satisfies the design temperature. demineralizer and return flow to the flow rate specified in Table 9.1.3-spent fuel pool. 1a AND b. The MCR indication for RPCS Place the RPCS in service to flow pump flow satisfies the design through a different pool cleanup flow rate specified in Table 9.1.3-filter and demineralizer and return 1b flow to the reactor pool. ii. a. SFPCS flow and RPCS flow to the ii. Simulate a high water temperature pool cleanup filters and upstream of one of the pool cleanup demineralizers stop.

filters. b. The SFPCS flow is bypassed to the spent fuel pool.

c. The RPCS cooling flow is bypassed to the reactor pool.

d. The MCR indication for SFPCS pump flow satisfies the design flow rate specified in Table 9.1.3-1a

e. The MCR indication for RPCS pump flow satisfies the design flow rate specified in Table 9.1.3-1b Tier 2 14.2-22 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-3: Reactor Pool Cooling System Test # 3 Preoperational test is required to be performed once.

The RPCS is described in Section 9.1.3.2.2. RPC system functions are not verified by RPCS tests. RPCS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

The RPCS supports the PCUS by nonsafety-related PCU Test #2-1 providing reactor pool water for purification of the UHS.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each RPCS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

ii. Verify each RPCS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each RPCS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each RPCS pump can be Align the RPCS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

v. Verify each RPCS pump automatically Align the RPCS to allow for pump MCR display and local, visual stops to protect the pump. operation. Place a pump in service. observation indicate each pump stops.

Initiate a simulated stop signal for the following system conditions.

i. Low pump suction pressure.

ii. High pump discharge pressure.

vi. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from an RPCS grab sample through the grab sampling device. obtained.

device indicated on the RPCS piping and instrumentation diagram.

vii. Verify each RPCS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each RPCS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-23 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-4: Pool Surge Control System Test #4 Preoperational test is required to be performed once.

The pool surge control system (PSCS) is described in Section 9.1.3.2.4 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The PSCS supports the UHS by providing nonsafety-related Test #4-1 surge control for UHS operations.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each PSCS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

ii. Verify each PSCS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each PSCS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each PSCS pump can be Stop and start each pump from the MCR. MCR display and local, visual started and stopped remotely. observation indicate each pump starts and stops.

v. Verify the PSCS automatically Initiate a real or simulated high radiation i. The PSCS tank inlet isolation valve is responds to mitigate a release of signal in the PSCS tank vent line. closed.

radioactivity. ii. The PSCS tank outlet isolation valve is closed.

[ITAAC 03.09.10]

vi. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a PSCS grab sample through the grab sampling device. obtained.

device indicated on the PSC piping and instrumentation diagram.

vii. Verify each PSC system instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each PSCS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS). historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-24 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-4: Pool Surge Control System Test #4 (Continued)

System Level Test #4-1 Test Objective Test Method Acceptance Criteria Verify PSCS automatic control for dry Align the PSCS for fill and drain of the dry i. a. Pump is stopped and return line dock fill and drain. dock. to pool surge control tank Fill the dry dock to a level that allows isolation valve is closed.

operation of the reactor inspection dry b. Pump is stopped and return line dock evacuation pump. to PSCS tank isolation valve is

i. Start a PSCS pump and simulate the closed.

following PSCS conditions: c. PSCS tank main discharge line

a. Dry dock low level isolation valve is closed.

b. PSC tank high level

c. High dry dock level Tier 2 14.2-25 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-5: Ultimate Heat Sink Test # 5 There are no preoperational tests for the UHS.

The UHS is described in Section 9.2.5. The only active functions for the UHS are to provide PAM Type D instrument signals to the safety display and indication system (SDIS). Refer to Table 14.2-66: Safety Display and Indication test

66 for testing of PAM Type D displays.

System Function System Function Categorization Function Verified by Test #

None N/A N/A Prerequisites:

N/A Component Level Tests None Tier 2 14.2-26 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-6: Pool Leak Detection System Test # 6 There are no preoperational tests for the pool leakage detection system (PLDS).

The PLDS is described in Section 9.1.3.2.5 Leakage from the UHS liner gravity drains to the radiation waste drain system (RWDS). Test #23-2 tests the MCR alarm when the RWDS sump fill rate exceeds the PLDS leakage rate setpoint.

System Function System Function Categorization Function Verified by Test #

None N/A N/A Prerequisites:

N/A Component Level Tests None Tier 2 14.2-27 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-7: Reactor Component Cooling Water System Test # 7 Preoperational test is required to be performed once for shared/common components, 6 times for the module-specific components on the 6A NPMs and once for shared/common components, 6 times for the module-specific components on the 6B NPMs.

The RCCWS is described in Section 9.2.2 and 11.5.2.2.12 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The reactor component cooling water nonsafety-related Test #7-1 system (RCCWS) supports the following systems by providing cooling water.

control rod drive system (CRDS)

chemical and volume control system (CVCS)

containment evacuation system (CES)

process sampling system (PSS)

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify an RCCWS flow balance has been performed.

iii. Verify a pump curve test has been completed for the RCCWS pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each RCCWS remotely- Operate each valve from the MCR and MCR display and local, visual operated valve can be operated local control panel (if design has local observation indicate each valve fully remotely. valve control) opens and fully closes.

ii. Verify each RCCWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each RCCWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each RCCWS pump can be Align the RCCWS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

v. Verify the RCCWS standby pump Align the RCCWS to allow for pump MCR display and local, visual automatically starts to protect plant operation. Place a pump in service. observation indicate the standby pump equipment. Initiate a simulated RCCWS pump low starts.

header pressure signal. Audible and visible water hammer are not observed when the pump starts.

vi. Verify RCCWS demineralized makeup i. Initiate simulated expansion tank MCR display and local, visual water level control valve high level signal. observation indicate the following:

automatically operates to maintain ii. Initiate a simulated expansion tank i. The demineralized makeup water RCCW expansion tank level. low level signal. level control valve is fully closed.

ii. The demineralized makeup water level control valve is fully open.

vii. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from an RCCWS grab through the grab sampling device. obtained.

sample device.

Tier 2 14.2-28 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-7: Reactor Component Cooling Water System Test # 7 (Continued) viii. Verify each RCCWS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each RCCWS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #7-1 Test Objective Test Method Acceptance Criteria Verify RCCWS cooling water flow rates Module 1 Test The RCCWS cooling flow to each heat satisfy design flow. i. Align the 6A RCCWS to provide flow exchanger under test meets the flow to all the Module 1 heat exchangers rate acceptance criteria contained in the cooled by RCCWS listed below: RCCWS flow balance report.

Module 1 Heat Exchangers Control rod drive mechanism (CRDM) cooling coils CVCS non-regenerative heat exchanger CES vacuum pump CES condenser PSS analyzer cooler PSS temperature control unit ii. Repeat module 1 test for modules 2 through 6.

iii. Align the 6B RCCWS to provide flow to all the Module 7 heat exchangers cooled by RCCWS listed below:

Module 7 Heat Exchangers CRDM cooling coils CVCS non-regenerative heat exchanger CES vacuum pump CES condenser PSS analyzer cooler PSS temperature control unit iv. Repeat Module 7 test for modules 8 through 12.

Tier 2 14.2-29 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-8: Chilled Water System Test # 8 Preoperational test is required to be performed once.

The chilled water system (CHWS) is described in Section 9.2.8 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The CHWS supports the following nonsafety-related Test #8-1 systems by providing cooling water: Test #8-2

Reactor Building HVAC system (RBVS)

normal control room HVAC system (CRVS)

Radioactive Waste Building HVAC system (RWBVS)

liquid radioactive waste system (LRWS)

gaseous radioactive waste system (GRWS)

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify a CHWS flow balance has been performed.

iii. Verify a pump curve test has been completed for the CHWS pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each CHWS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

ii. Verify each CHWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each CHWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify the speed of each CHWS Align the CHWS to provide a flow path to MCR display indicates the speed of each variable-speed pump can be operate a selected pump. obtains both minimum and maximum manually controlled. Vary the CHWS pump speed from pump speeds.

minimum to maximum from the MCR. Audible and visible water hammer are not observed when the pump starts.

v. Verify automatic operation of CHWS Align the CHWS to allow for chiller MCR display and local, visual pumps and CHWS chiller to protect operation. Place a pump in service. observation indicate the following:

plant equipment. Initiate a simulated start signal for the i. a. Operating pump stops following system conditions. b. Operating chiller stops

i. Loss of chilled water flow. ii. Operating chiller stops ii. Loss of SCWS cooling flow to the operating chiller.

Tier 2 14.2-30 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-8: Chilled Water System Test # 8 (Continued) vi. Verify each CHWS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each CHWS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #8-1 Test Objective Test Method Acceptance Criteria Verify CHWS cooling water flow rates i. Align the CHWS to provide flow to all The CHWS cooling flow to each heat satisfy design. heat exchangers cooled by the CRVS exchanger under test meets the chiller: minimum flow rate acceptance criteria RBVS air handling units contained in the CHWS flow balance RBVS fan coil units report.

CRVS air handling units CRVS fan coil units RWBVS air handling units RWBVS fan coil units LRW degasifier condenser GRWS gas coolers ii. Open all CHWS flow control valves.

System Level Test #8-2 Test Objective Test Method Acceptance Criteria Verify CHWS cooling water flow rates i. Align the CHWS to provide flow to The CRVS standby CHWS cooling flow to satisfy design flow. the CRVS air handling units and the each heat exchanger meets the CRVS fan coil units cooled by the minimum flow rate acceptance criteria CRVS standby chiller. contained in the CHWS flow balance ii. Open all CHWS flow control valves. report.

Tier 2 14.2-31 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-9: Auxiliary Boiler System Test # 9 Preoperational test is required to be performed once.

The auxiliary boiler system (ABS) is described in Section 10.4.10 and 11.5.2.2.14 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

None The ABS functions verified by other tests are:

The auxiliary boiler supports the nonsafety-related CPS Test #30-1 condensate polishing system (CPS) by supplying steam for resin regeneration.

The auxiliary boiler supports the turbine nonsafety-related CAR Test #32-1 generator by supplying gland seal steam.

The auxiliary boiler supports the FWS by nonsafety-related CAR Test #32-1 supplying steam to the condenser for sparging when necessary.

The auxiliary boiler supports the module nonsafety-related TG Test #33-1 heatup system (MHS) by supplying steam for heating reactor coolant at startup and shutdown.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify a pump curve test has been completed for the auxiliary boiler pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each auxiliary boiler remotely- Operate each valve from the MCR and MCR display and local, visual operated valve can be operated local control panel (if design has local observation indicate each valve fully remotely. valve control) opens and fully closes.

ii. Verify each auxiliary boiler air- Place each valve in its non-safe position. MCR display and local, visual operated valve fails to its safe Isolate and vent air to the valve. observation indicate each valve fails to position on loss of air. its safe position.

iii. Verify each auxiliary boiler air- Place each valve in its non-safe position. MCR display and local, visual operated valve fails to its safe Isolate electrical power to each air- observation indicate each valve fails to position on loss of electrical power to operated valve. its safe position.

its solenoid.

iv. Verify each auxiliary boiler low Align the ABS to allow for pump MCR display and local, visual pressure boiler feedwater pump can operation. observation indicate each pump starts be started and stopped remotely. Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

v. Verify the speed of each auxiliary Align the ABS to provide a flow path to MCR display indicates the speed of each boiler high pressure boiler feedwater operate a selected AB variable-speed variable speed pump obtains both pump can be manually controlled. pump. minimum and maximum pump speeds.

Vary the auxiliary boiler pump speed Audible and visible water hammer are from minimum to maximum speed from not observed when the pump starts.

the MCR.

vi. Verify the ABS automatically Initiate a real or simulated high radiation MCR display verifies the following:

responds to mitigate a release of signal for the auxiliary boiler flash tank i. auxiliary boiler flash tank vent radioactivity. vent. isolation valve is closed.

ii. auxiliary boiler high pressure steam supply isolation valves are closed.

[ITAAC 03.09.08]

(i.and ii.)

Tier 2 14.2-32 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-9: Auxiliary Boiler System Test # 9 (Continued) vii. Verify the ABS automatically Initiate a real or simulated high radiation MCR display verifies the following:

responds to mitigate a release of signal for the auxiliary boiler high auxiliary boiler high pressure to low radioactivity. pressure to low pressure steam supply. pressure steam supply pressure control valve is closed.

[ITAAC 03.09.09]

viii. Verify each ABS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each ABS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test Test Objective Test Method Acceptance Criteria None Tier 2 14.2-33 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-10: Circulating Water System Test # 10 This preoperational test is required to be performed once for each circulating water subsystem.

The circulating water system (CWS) is described in Section 10.4.5 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The utility water system (UWS) supports nonsafety-related Component-Level Test vi.

the CWS by providing makeup water to maintain water level in the CWS cooling tower basins.

The CWS function verified by another test is:

System Function System Function Categorization Function Verified by Test #

The CWS supports the FWS by removing nonsafety-related CAR Test #32-1 heat from the main condenser.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests: NPM #1 (#7)

The minimum inventory of pumps, fans and valves tested for NPM #1 (#7) is that inventory required for 6A (6B) CWS operation to support operation of NPM #1 (#7). The testing will continue until all 6A (6B) CWS equipment is tested.

Test Objective Test Method Acceptance Criteria

i. Verify each CWS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

ii. Verify each CWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each CWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each CWS cooling tower fan Align the CWS to allow for cooling tower MCR display and local, visual can be started and stopped remotely fan operation. observation indicate each cooling tower Stop and start each cooling tower fan fan starts and stops.

from the MCR.

v. Verify each CWS pump can be started Align the CWS to allow for pump i. MCR display and local, visual and stopped remotely. operation. observation indicate each pump Stop and start each pump from the MCR. starts and stops.

ii. Audible and visible water hammer are not observed when the pump starts.

iii. CWS pump cavitation is not observed.

iv. Cooling towers do not experience flow surge or overflow.

vi. Verify automatic operation of the i. Initiate a cooling tower basin low MCR displays and local, visual CWS cooling tower basin level level signal. observation verifies the following:

control valve to maintain CWS ii. Initiate a cooling tower basin high i. The cooling tower basin level cooling tower basin level. level signal. control valve is open.

ii. The cooling tower basin level control valve is closed.

Tier 2 14.2-34 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-10: Circulating Water System Test # 10 (Continued) vii. Verify each CWS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each CWS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-35 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-11: Site Cooling Water System Test # 11 Preoperational test is required to be performed for each NPM.

The site cooling water system (SCWS) is described in Section 9.2.7 and 11.5.2.2.13 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

The SCWS supports the following nonsafety-related Test #11-1 systems by providing cooling water.

turbine generator system (TGS)

RCCWS

condenser air removal system (CARS)

PSS

CHWS

instrument air system (IAS)

SFPCS

RPCS

auxiliary boiler The UWS supports the SCWS by nonsafety-related Component-Level Test vii.

providing makeup water to maintain water level in the SCWS cooling tower basins.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify an SCWS flow balance has been performed and the system flow balance records have been approved.

iii. Verify a pump curve test has been completed and approved for the SCWS pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each SCWS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each SCWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each SCWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each SCWS cooling tower fan Align the SCWS to allow for cooling MCR display and local, visual can be started and stopped remotely. tower fan operation. observation indicate each cooling tower Stop and start each cooling tower fan fan starts and stops.

from the MCR.

v. Verify each SCWS pump can be Align the SCWS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

vi. Verify the SCWS standby pump Align the SCWS to allow for pump MCR display and local, visual automatically starts to protect plant operation. Place a pump in service. observation indicate the standby pump equipment. Initiate a simulated start signal for the discharge valve opens to a throttled following system conditions. position, the pump starts, and then the

i. Low pump header pressure signal. discharge valve fully opens.

ii. Low pump header flow signal. Audible and visible water hammer are not observed when the pump starts.

Tier 2 14.2-36 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-11: Site Cooling Water System Test # 11 (Continued) vii. Verify automatic operation of the i. Initiate a simulated cooling tower MCR displays and local, visual SCWS cooling tower basin level basin low level signal. observation verifies the following:

control valve to maintain SCWS ii. Initiate a simulated cooling tower i. The cooling tower basin level cooling tower basin level. basin high level signal. control valve is open.

ii. The cooling tower basin level control valve is closed.

viii. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a SCWS grab sample through the grab sampling device. obtained.

device indicated on the SCWS piping and instrumentation diagram.

ix. Verify each SCWS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each SCWS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-37 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-11: Site Cooling Water System Test # 11 (Continued)

System Level Test #11-1 Test Objective Test Method Acceptance Criteria Verify SCWS cooling water flow rates i. NPM 1 Test The SCWS cooling flow to each heat satisfy design flow. Align the SCWS to provide flow to all exchanger under test meets the the Module 1 and common heat minimum flow rate acceptance criteria exchangers cooled by SCWS listed contained in the SCWS flow balance below: report.

NPM 1 Heat Exchangers CARS heat exchanger TGS cooler TGS lube oil cooler TGS DEHC cooler FWS sample cooler MSS sample cooler Common Heat Exchangers CHWS chillers Instrument air coolers PSS chillers RPCS heat exchangers SFPCS heat exchangers RCCWS heat exchangers Aux boiler blowdown coolers The operation of two SCWS pumps may be required to provide sufficient flow to meet acceptance criteria in the SCWS flow balance report.

ii. Reactor Power Module 2-12 Test The scope of each subsequent test will include one or more additional modules. The scope will also include previously tested modules to verify that the flow rate still meets the flow rate acceptance criteria contained in the SCWS flow balance report. The testing will continue until all heat exchangers cooled by SCWS have been tested in a single test.

Tier 2 14.2-38 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-12: Potable Water System Test # 12 The potable water system (PWS) is described in Section 9.2.4. The PWS is a site-specific system, and the testing of the PWS is the responsibility of the COL applicant.

COL Item 14.2-5: A COL applicant that references the NuScale Power Plant design certification will provide a test abstract for the potable water system pre-operational testing.

System Function System Function Categorization Function Verified by Test #

As described in Section 9.2.4 nonsafety-related Provided by COL applicant Prerequisites Provided by COL applicant Component Level Tests Provided by COL applicant System Level Tests Provided by COL applicant Tier 2 14.2-39 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-13: Utility Water System Test # 13 Preoperational test is required to be performed once.

The UWS is described in Section 9.2.9 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The UWS supports the circulating nonsafety-related Reference 14.2-10 Component-Level water system by providing makeup Test vi.

water to maintain water level in the CW system cooling tower basins.

2. The UWS supports the SCWS by nonsafety related Reference 14.2-11 Component-Level providing makeup water to maintain Test vii.

water level in the SCWS cooling tower basins.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify a pump curve test has been completed for the UWS pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each UWS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each UWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each UWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each UWS pump can be started Align the UWS to allow for pump MCR display and local, visual and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

v. Verify UWS flow capability by Align the UWS to allow for pump MCR display and local, visual automatic start of each UWS pump operation. Place a pump in service. observation indicate the standby pump while in standby mode. Initiate a simulated pump discharge starts.

pressure lowl. Audible and visible water hammer are not observed when the pump starts.

vi. Verify demineralized water system Align the DWS to allow for pump MCR display and local, visual (DWS) pumps automatically stops to operation. Place a pump in service. observation indicate each pump stops.

protect plant equipment. Initiate a simulated DWS storage tank low level signal.

vii. Verify pump low flow protection i. Align the DWS to allow for DWS MCR displays and local, visual pump operation. Place a DWS pump observation verifies the following:

in operation. i. The pump minimum flow valve is Manually throttle a valve in the DWS open.

pump flow path until the pump flow ii. The pump minimum flow valve is rate reaches the pump minimum closed.

flow setpoint.

ii. Open the throttled valve.

Tier 2 14.2-40 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-13: Utility Water System Test # 13 (Continued) viii. Verify each DWS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each DWS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-41 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-14: Demineralized Water System Test # 14 Preoperational test is required to be performed once.

The DWS is described in Section 9.2.3 and 11.5.2.2.16 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The DWS supports the following systems nonsafety-related component-level tests by providing cooling water.

CVCSm

boron addition system (BAS)

Liquid Radioactive Waste (LRW)

SFPCS

RCCWS

Process Sampling System (PSS)

FWS

ABS

CRVS

CARS

CES Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify a pump curve test has been completed for the DWS pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each DWS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

ii. Verify each DWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each DWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify the DWS pump can be started Align the DWS to allow for pump MCR display and local, visual and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

v. Verify DWS flow capability by Align the DWS to allow for pump MCR display and local, visual automatic start of each DWS pump operation. Place a pump in service. observation indicate the standby pump while in standby mode. Initiate a simulated low pump header starts.

pressure low signal. Audible and visible water hammer are not observed when the pump starts.

vi. Verify DWS pumps automatically Align the DWS to allow for pump MCR display and local, visual stops to protect plant equipment. operation. Place a pump in service. observation indicate each pump stops.

Initiate a simulated DWS storage tank low level signal.

vii. Verify pump low flow protection i. Align the DWS to allow for DWS MCR displays and local, visual pump operation. Place a DWS pump observation verifies the following:

in operation. i. The pump minimum flow valve is Manually throttle a valve in the DWS open.

pump flow path until the pump flow ii. The pump minimum flow valve is rate reaches the pump minimum closed.

flow setpoint.

ii. Open the throttled valve.

Tier 2 14.2-42 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-14: Demineralized Water System Test # 14 (Continued) viii. Verify each DWS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each DWS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-43 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-15: Nitrogen Distribution System Test # 15 Preoperational test is required to be performed once.

The nitrogen distribution system (NDS) is described in Section 9.3.1 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

Has no specific system function, all N/A N/A functionality is supported through supported systems testing.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each NDS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

ii. Verify each NDS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each NDS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify the NDS high pressure Initiate a simulated signal for the MCR display and local, visual isolation valve closes to protect following system conditions. observation indicate the nitrogen equipment. i. High flow to high pressure header supply to the high pressure header valve ii. High pressure on high pressure is closed.

header

v. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a NDS grab sample through the grab sampling device. obtained.

device indicated on the NDS piping and instrumentation diagram.

vi. Verify each NDS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each NDS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-44 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-16: Service Air System Test # 16 Preoperational test is required to be performed once.

The service air system (SAS) is described in Section 9.3.1 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

Has no specific system function, all N/A N/A functionality is supported through supported systems testing.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each SAS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

ii. Verify each SAS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each SAS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify the instrument air to service air Initiate a simulated signal for the MCR display and local, visual isolation valve closes to protect following system conditions. observation indicate the instrument air equipment. i. IAS header low pressure to service air isolation valve closes.

ii. SAS high flow iii. SAS low pressure iv. SAS filter high differential pressure

v. Verify each SAS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each SAS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-45 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-17: Instrument Air System Test # 17 Preoperational test is required to be performed once.

The IAS is described in Section 9.3.1 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

Has no specific system function, all N/A N/A functionality is supported through supported systems testing.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test. Verify performance testing of air compressor skids have been completed by the manufacturer or a site acceptance test has been completed in accordance with manufacturer instructions.

Component Level Tests: First NPM Test Objective Test Method Acceptance Criteria

i. Verify each IAS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each IAS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each IAS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify the speed of each IAS variable- Align the system to provide a flow path MCR display indicate the speed of each speed compressor can be manually to operate a selected compressor. compressor obtains both minimum and controlled. Vary the compressor speed from maximum pump speeds.

minimum to maximum from the MCR.

v. Verify each IAS standby compressor Align the SCWS to allow for compressor MCR display and local, visual automatically starts to protect plant operation. Place a compressor in service. observation indicate the following equipment. Initiate a real or simulated start signal for standby compressor starts.

receiver low pressure.

vi. Verify each IAS operating compressor Align the IAS to allow for compressor MCR display and local, visual automatically stops to protect the operation. Place a compressor in service. observation indicate the operating compressor. Initiate a simulated signal for the compressor stops.

following system conditions.

i. Pre-filter high differential pressure protection ii. Post-filter high differential pressure protection iii. High dew point protection Tier 2 14.2-46 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-17: Instrument Air System Test # 17 (Continued) vii. Verify each IAS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each IAS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Component Level Tests: NPMs 2-12 Test Objective Test Method Acceptance Criteria

i. Verify each IAS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

ii. Verify each IAS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each IAS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

System Level Tests None Tier 2 14.2-47 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-18: Control Room Habitability System Test # 18 Preoperational test is required to be performed once.

The control room habitability system (CRHS) is described in Section 6.4 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The CRHS supports the Control nonsafety-related Test #18-1 Building (CRB) by providing clean Test #18-2 breathing air to the control room envelope (CRE) and maintaining a positive control room pressure during high radiation or loss of offsite power conditions.

2. The CRHS supports the CRB by nonsafety-related Test #18-1 providing high pressure, clean Test #18-2 breathing air in air bottles for use.

3. The CRVS supports the CRB by nonsafety-related Test #18-1 providing isolation of the CRE from the surrounding areas and outside environment via isolation dampers.

4. The plant protection system (PPS) nonsafety-related Test #18-1 supports the CRHS by providing actuation and control signals.

5. The CRVS supports the CRB by nonsafety-related Test #18-1 providing isolation of the CRE from the surrounding areas and outside environment via isolation dampers.

6. The CRVS supports the PPS by nonsafety-related Test #18-1 providing instrument information signals relating to isolation of the CRE and activation of the CRH system.

7. The CRVS supports the CRB by nonsafety-related Test #18-1 isolating the CRVS outside air intake when radiation is detected downstream of the charcoal filtration unit.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify a CRHS air balance has been performed and the CRHS air balance records have been approved. [This prerequisite is not required for component-level tests.]

iii. Verify CRHS air bottlers are pressurized to their design working pressure. [This prerequisite is not required for component-level tests.]

Tier 2 14.2-48 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-18: Control Room Habitability System Test # 18 (Continued)

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each CRHS remotely-operated Place the CRHS air bottles in service. i. MCR workstation display, safety valve can be operated remotely. Place CRVS in service to supply air to the display instrument display and local, CRE. visual observation indicate each Operate each valve from the MCR. valve fully opens and fully closes under preoperational temperature, differential pressure, and flow conditions.

[ITAAC 03.01.02]

ii. Verify each CRHS solenoid-operated Place the CRHS air bottles in service. i. MCR display, safety display valve fails to its safe position on loss Place CRVS in service to supply air to the instrument display and local, visual of electrical power to its solenoid. CRE. observation indicate each valve fails

i. Place each valve in its non-safe open under preoperational position. Isolate electrical power to temperature, differential pressure, its solenoid. and flow conditions.

[ITAAC 03.01.03]

iii. Verify each CRHS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each CRHS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #18-1 Test Objective Test Method Acceptance Criteria Verify the CRHS and the CRVS Place the CRVS in automatic operation. MCR workstation display and local, automatically respond to provide Place the CRHS air bottles in service. visual observation indicate the breathable air to the CRE under accident Place CRVS in service to supply air to the following:

conditions. CRE. i. The CRVS outside air damper closes.

Initiate each of the following real or ii. The CRVS filter unit fan stops.

simulated CRHS actuation signals: iii. The CRVS control room envelope

High radiation signal downstream of isolation dampers close.

the CRVS filter unit iv. The CRHS air supply isolation valves

Loss of AC power. open.

v. CRHS pressure relief isolation valves open.

vi. CRVS air handling unit stops.

vii. CRE general exhaust fan stops.

viii. CRVS battery room exhaust fan stops.

[ITAAC 03.09.02]

(items i through v)

Tier 2 14.2-49 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-18: Control Room Habitability System Test # 18 (Continued)

System Level Test #18-2 Test Objective Test Method Acceptance Criteria Verify emergency pressurized air bottles i. Align air bottles for testing. Assume i. a. The CRE described in Section have sufficient volume to provide 72 25% of the bottles are unavailable 6.4.2.1 maintains a positive hours of breathable air through both the and use 1/6 of the remaining bottles pressure relative to the adjacent main and backup supply flow path to the to simulate a test conduct of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months areas as specified in Table 6.4-1 CRE described in Section 6.4.2.1. (12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s/72 hours). as indicated by the CRE Initiate a real or simulated CRHS differential pressure actuation signal to isolate the CRE. transmitters.

Conduct a CRE test for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . [ITAAC 03.01.05]

ii. At the end of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months isolate the b. The CRHS minimum flow rate main supply flow path and align the for the main flow path is manual backup flow path to the CRE. maintained as specified in Table Align air bottles for testing. Assume 6.4-1 for the duration of the test.

25% of the bottles are unavailable c. The CRHS flow rate for the and use the remaining bottles. manual backup flow path is maintained as specified in Table 6.4-1.

System Level Test #18-3 Test Objective Test Method Acceptance Criteria The air exfiltration from the CRE does not Perform an air exfiltration test of the CRE The measured air exfiltration flow rate exceed the air exfiltration flow rate at 1/8 in. wg. of positive pressure with does not exceed the unfiltered identified in the CRHS exfiltration/ respect to surrounding areas by inleakage flow rate assumed in the dose infiltration analysis. performing tracer gas testing in analysis identified in Table 6.4-1.

accordance with ASTM E741. [ITAAC 03.01.01]

Tier 2 14.2-50 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-19: Normal Control Room HVAC System Test # 19 Preoperational test is required to be performed once.

The CRVS is described in Sections 6.4.3.2, 9.4.1, and 11.5.2.2.1, and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The CRVS supports the CRB by nonsafety-related Test #19-1 providing cooling, heating and Test #19-2 humidity control to maintain a suitable environment for the safety and comfort of plant personnel.

2. The CRVS supports the systems nonsafety-related Test #19-1 located in the CRB by providing Test #19-2 cooling, heating and humidity control to maintain a suitable environment for the operation of system components.

3. The CRVS supports the CRB by nonsafety-related Test #19-1 maintaining the CRB at a positive pressure with respect to adjacent areas during normal operation.

4. The CRVS supports the CRB by nonsafety-related Test #19-3 maintaining the CRB at a positive ambient pressure relative to the Reactor Building (RXB) and the outside atmosphere to control the ingress of potentially airborne radioactivity from the RXB or the outside atmosphere to the CRB.

5. The PPS supports the CRVS by nonsafety-related Test #19-3 providing actuation and control signals to the outside air isolation dampers.

6. The CRVS supports the CRB by nonsafety-related Test #19-4 protecting personnel from exposure to radiation during a design basis accident, when power is available, by removing radioactive contamination from outside air via charcoal filtration, as required by radiation dose analyses.

The CRVS functions verified by other tests are:

The CRVS supports the CRB by isolating nonsafety-related Test #18-1 the CRVS outside air intake when radiation is detected downstream of the charcoal filtration unit.

The CRVS supports the CRB by providing nonsafety-related Test #18-1 isolation of the CRE from the surrounding areas and outside environment via isolation dampers.

The CRVS supports the PPS by providing nonsafety-related Test #18-1 instrument information signals relating to isolation of the CRE and activation of the CRHS.

Tier 2 14.2-51 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-19: Normal Control Room HVAC System Test # 19 (Continued)

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify a CRVS air balance has been performed and the CRVS air balance records have been approved. [This prerequisite is not required for component-level tests.]

iii. Verify CRVS high-efficiency particulate air (HEPA) and charcoal adsorbers have been installed and tested and the test records have been approved. [This prerequisite is not required for component-level tests.]

iv. Verify CRVS control room isolation dampers have been leak tested and the test records have been approved. [This prerequisite is not required for component-level tests.]

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each CRVS remotely-operated Operate each damper from the MCR and MCR display and local, visual damper can be operated remotely. local control panel (if design has local observation indicate each damper fully damper control). opens and fully closes.

ii. Verify each CRVS air-operated Place each damper in its non-safe MCR display and local, visual damper fails to its safe position on position. Isolate and vent air to the observation indicate each damper fails loss of air. damper. to its safe position.

iii. Verify each CRVS air-operated Place each damper in its non-safe MCR display and local, visual damper fails to its safe position on position. Isolate electrical power to its observation indicate each damper fails loss of electrical power to its solenoid. to its safe position.

solenoid.

iv. Verify CRVS dampers automatically Open each damper actuated by a smoke MCR display and local, visual close on associated smoke or fire or fire signal. Initiate an alarm signal for observation indicate each damper signals. each damper. closes.

v. Verify each required CRVS fan stops Initiate an alarm signal for each fan. MCR display and local, visual on actuation of its associated fire or observation indicate each fan stops.

smoke alarm.

vi. Verify each CRVS pressurization fan Initiate an alarm signal for each fan. MCR display and local, visual starts automatically on the actuation observation indicate each pressurization of its associated fire or smoke alarm. fan starts.

vii. Verify the fan speed of each CRVS Vary the speed of each fan from the MCR MCR display indicates the speed of each variable-speed fan can be manually and local control panel (if design has fan varies from minimum to maximum controlled. local fan control). speed.

viii. Verify the standby CRVS main supply Place an AHU in service. Place the MCR display and local, visual air handling unit (AHU) starts standby AHU in automatic control. Stop observation indicate the standby AHU automatically on the stop of the the operating AHU. starts.

operating CRVS main supply AHU.

ix. Verify each standby CRVS fan coil unit Place an FCU in service. Place the standby MCR display and local, visual (FCU) starts automatically on the stop FCU in automatic control. Stop the observation indicate the standby FCU of the operating CRVS fan coil unit. operating FCU. starts.

x. Verify each CRVS control room Place each damper in its non-safe Each CRVS control room envelope envelope isolation damper fails to its position. Isolate and vent air to the isolation damper fails to its closed safe position on loss of air. damper. position on loss of air under preoperational temperature, differential pressure, and flow conditions while the CRV system is supplying flow to the CRE.

[ITAAC 03.02.01]

xi. Verify each CRVS control room Place each damper in its non-safe Each CRVS control room envelope envelope isolation damper fails to its position. Isolate electrical power to its isolation damper fails to its closed safe position on loss of electrical solenoid. position on loss of electrical power power to its solenoid. under preoperational temperature, differential pressure, and flow conditions while the CRVS is supplying flow to the CRE.

[ITAAC 03.02.01]

Tier 2 14.2-52 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-19: Normal Control Room HVAC System Test # 19 (Continued) xii. Verify each CRVS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each CRVS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #19-1 Test Objective Test Method Acceptance Criteria

i. Verify CRB design temperatures and Place the CRVS in automatic operation. i. The temperature and humidity of humidity monitored by the MCR are i. Record the CRB temperatures and rooms and areas monitored by the maintained at design temperature humidity indications monitored by MCR satisfy the design temperature and humidity conditions during the MCR. and humidity requirements normal operation. ii. Measure the CRB pressure relative contained in Table 9.4.1-2.

ii. Verify The CRVS maintains a positive the outside environment. ii. The CRVS maintains a positive pressure in the CRB relative to the iii. Measure the air flow rate to the pressure of greater than or equal to outside environment while the CRVS battery rooms. 0.125 inches water gauge in the CRB is operating in normal alignment. relative to the outside environment, iii. Verify the CRVS maintains the air flow while operating in the normal to the battery rooms to maintain operating alignment.

hydrogen concentration to less than [ITAAC 03.02.02]

1% by volume.

iii. Measured flow to the battery rooms is equal to or greater than the flow specified by the air flow balance.

[ITAAC 03.02.03]

System Level Test #19-2 Test Objective Test Method Acceptance Criteria

i. Verify CRB design temperatures and Align the CHWS standby chiller to cool The temperature and humidity of rooms humidity monitored by the MCR are each CRVS main supply AHU. and areas monitored by the MCR satisfy maintained at design temperature Place the CRVS in automatic operation. the design temperature and humidity and humidity conditions while requirements contained in Table 9.4.1-2.

cooling to the CRV main supply AHU is supplied by the CHWS standby chiller.

System Level Test #19-3 Test Objective Test Method Acceptance Criteria Verify the CRVS isolates makeup air when Place the CRVS in automatic operation. Outside air damper is closed to isolate smoke or toxic gas is detected in the i. Initiate a real or simulated high makeup air.

makeup air ductwork. radiation signal for the makeup air ductwork upstream of the CRVS filter unit.

Tier 2 14.2-53 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-19: Normal Control Room HVAC System Test # 19 (Continued)

System Level Test #19-4 Test Objective Test Method Acceptance Criteria Verify the CRVS automatically responds Place the CRVS in automatic operation. i. Outside air is diverted through the to mitigate the consequences of high Initiate a real or simulated high radiation CRVS filter unit by closing the CRVS radiation in the outside air. signal for the outside air ductwork filter unit bypass dampers and upstream of the CRVS filter unit. opening the CRVS filter unit isolation dampers.

ii. The CRVS filter unit fan starts.

[ITAAC 03.09.01]

(items i. and ii.)

Tier 2 14.2-54 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-20: Reactor Building HVAC System Test # 20 Preoperational test is required to be performed once.

The RBVS is described in Section 9.4.2 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The RBVS supports the RXB by nonsafety-related Test #20-1 providing cooling, heating and Test #20-2 humidity control to maintain a suitable environment for the safety and comfort of plant personnel.

2. The RBVS supports the systems nonsafety-related Test #20-1 located in the RXB by providing Test #20-2 cooling, heating and humidity control to maintain a suitable environment for the operation of system components.

3. The RBVS supports the RXB by nonsafety-related Test #20-1 maintaining the RXB at a negative Test #20-3 ambient pressure relative to the outside atmosphere to control the movement of potentially airborne radioactivity from the RXB to the environment.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify an RBVS air balance has been performed and the RBV system air balance records have been approved. [This prerequisite is not required for component-level tests.]

iii. RBVS high-efficiency particulate air and charcoal adsorbers have been installed and tested. [This prerequisite is not required for component-level tests.]

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each RBVS remotely-operated Operate each damper from the MCR and MCR display and local, visual damper can be operated remotely. local control panel (if design has local observation indicate each damper fully damper control). opens and fully closes.

ii. Verify each RBVS air-operated Place each damper in its non-safe MCR display and local, visual damper fails to its safe position on position. Isolate and vent air to the observation indicate each damper fails loss of air. damper. to its safe position.

iii. Verify each RBVS air-operated Place each damper in its non-safe MCR display and local, visual damper fails to its safe position on position. Isolate electrical power to its observation indicate each damper fails loss of electrical power to its solenoid. to its safe position.

solenoid.

iv. Verify RBVS dampers automatically Open each damper actuated by a smoke MCR display and local, visual close on associated smoke or fire or fire signal. Initiate an alarm signal for observation indicate each damper signals. each damper. closes.

v. Verify each required RBVS fan stops Initiate an alarm signal for each fan. MCR display and local, visual on actuation of its associated fire or observation indicate each fan stops.

smoke alarm.

vi. Verify each RBVS pressurization fan Initiate an alarm signal for each fan. MCR display and local, visual starts automatically on the actuation observation indicate each RWBVS of its associated fire or smoke alarm. pressurization fan starts.

vii. Verify the fan speed of each RBVS Vary the speed of each fan from the MCR MCR display indicates the speed of each variable-speed fan can be manually and local control panel (if design has fan varies from minimum to maximum controlled. local fan control). speed.

Tier 2 14.2-55 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-20: Reactor Building HVAC System Test # 20 (Continued) viii. Verify each standby RBVS air Place an AHU in service. Place the MCR display and local, visual handling unit starts automatically on standby AHU in automatic control. Stop observation indicate the standby AHU the stop of the operating RBVS air the operating AHU. starts.

handling unit.

ix. Verify each standby RBVS fan coil unit Place an FCU in service. Place the standby MCR display and local, visual starts automatically on the stop of FCU in automatic control. Stop the observation indicate the standby FCU the operating RBVS fan coil unit. operating FCU. starts.

x. Verify each RBV system instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each RBVS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

xi. Verify each RBVS remotely-operated Operate each damper from the MCR and MCR display and local, visual damper can be operated remotely. local control panel (if design has local observation indicate each damper fully damper control) opens and fully closes.

Tier 2 14.2-56 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-20: Reactor Building HVAC System Test # 20 (Continued)

System Level Test #20-1 Test Objective Test Method Acceptance Criteria

i. Verify RXB design temperatures and Place the RBVS supply, general area i. The temperature and humidity of humidity monitored by the MCR are exhaust and spent fuel pool exhaust in rooms and areas monitored by the maintained at design temperature automatic operation. Place the RWBVS in MCR satisfy the design temperature and humidity conditions during automatic operation. and humidity requirements normal operation. i. Record the RXB temperatures and contained in Table 9.4.2-2.

ii. Verify The RBVS maintains a negative humidity indications monitored by ii. MCR display indicates the RBVS pressure in the RXB relative to the the MCR. maintains a negative pressure in the outside environment while the RBVS ii. Measure the RXB pressure relative RXB relative to the outside is operating in normal alignment. the outside environment. environment while operating in the iii. Verify The RBVS maintains a negative iii. Measure the RWB pressure relative normal operating alignment.

pressure in the Radioactive Waste the outside environment. [ITAAC 03.03.01]

Building (RWB) relative to the outside iv. Measure the air flow rate to the environment while the RBVS is battery rooms. iii. MCR display indicates the RBVS operating in normal alignment. maintains a negative pressure in the iv. Verify the RBVS maintains the air flow RWB relative to the outside to the battery rooms to maintain environment while operating in the hydrogen concentration to less than normal operating alignment.

1% by volume. [ITAAC 03.03.02]

iii. Measured flow to the battery rooms is equal to or greater than the flow specified by the air flow balance.

[ITAAC 03.03.03]

System Level Test #20-2 Test Objective Test Method Acceptance Criteria

i. Verify design temperatures of the Place the RBVS air handling units with i. The temperature and humidity of following rooms can be controlled installed direct expansion coils in rooms and areas monitored by the using AHUs with installed direct automatic operation. MCR satisfy the design temperature expansion coils. and humidity requirements

a. RSS contained in Table 9.4.2-2.

b. module protection system (MPS) equipment rooms

c. battery rooms

d. battery charger rooms Tier 2 14.2-57 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-20: Reactor Building HVAC System Test # 20 (Continued)

System Level Test #20-3 Test Objective Test Method Acceptance Criteria

i. Verify RBVS automatic alignment on Place the RBVS general area exhaust, i. The RBVS general area exhaust a simulated spent fuel pool hi-hi RBVS spent fuel pool exhaust, RWBVS isolation damper for the spent fuel radiation level. exhaust and Annex Building (ANB) pool and dry dock area is closed to ii. Verify The RBVS maintains a negative exhaust in automatic operation. isolate the spent fuel pool area pressure in the RXB relative to the Place the RBVS supply in automatic exhaust flow from the RBVS general outside environment while the RBVS operation. exhaust.

is operating in accident alignment. Place the RWBVS supply system in ii. The RBVS diverts spent fuel pool iii. Verify The RWBVS maintains a automatic operation. exhaust flow to charcoal adsorbers negative pressure in the RWB relative Simulate a Hi-Hi radiation signal in the and additional HEPAs in the spent to the outside environment while the spent fuel pool exhaust upstream of the fuel pool charcoal filter units.

RBVS is operating in accident spent fuel pool charcoal filter units. iii. Flow from the RBVS supply fans is alignment. reduced to maintain the design negative pressure in the RXB and RWB relative to the outside environment while the RBVS is operating in the off- normal alignment.

[ITAAC 03.09.03]

(items i thru iii)

Tier 2 14.2-58 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-21: Radioactive Waste Building HVAC System Test # 21 Preoperational test is required to be performed once.

The RWBVS is described in Section 9.4.3 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The RWBVS supports the RWB by nonsafety-related Test #21-1 providing cooling, heating and humidity control to maintain a suitable environment for the safety and comfort of plant personnel.

2. The RWBVS supports the systems nonsafety-related Test #21-1 located in the RWB by providing cooling, heating and humidity control to maintain a suitable environment for the operation of system components.

3. The RWBVS supports the RWB by nonsafety-related Test #21-1 maintaining the RWB at a negative (normal RBVS exhaust alignment) ambient pressure relative to the outside atmosphere to control the Test #20-3 movement of potentially airborne radioactivity from the RWB to the (off-normal RBVS exhaust alignment) environment.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify an RWBVS air balance has been performed and the RWBVS air balance records have been approved. [This prerequisite is not required for component-level tests.]

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each RWBVS remotely- Operate each damper from the MCR and MCR display and local, visual operated damper can be operated local control panel (if design has local observation indicate each damper fully remotely. damper control). opens and fully closes.

ii. Verify each RWBVS air-operated Place each damper in its non-safe MCR display and local, visual damper fails to its safe position on position. Isolate and vent air to the observation indicate each damper fails loss of air. damper. to its safe position.

iii. Verify each RWBVS air-operated Place each damper in its non-safe MCR display and local, visual damper fails to its safe position on position. Isolate electrical power to its observation indicate each damper fails loss of electrical power to its solenoid. to its safe position.

solenoid.

iv. Verify RWBVS dampers automatically Open each damper actuated by a smoke MCR display and local, visual close on associated smoke or fire or fire signal. Initiate an alarm signal for observation indicate each damper signals. each damper. closes.

v. Verify each required RWBVS fan stops Initiate an alarm signal for each fan. MCR display and local, visual on actuation of its associated fire or observation indicate each fan stops.

smoke alarm.

vi. Verify each RWBVS pressurization fan Initiate an alarm signal for each fan. MCR display and local, visual starts automatically on the actuation observation indicate each pressurization of its associated fire or smoke alarm. fan starts.

vii. Verify the fan speed of each RWBVS Vary the speed of each fan from the MCR MCR display indicates the speed of each variable-speed fan can be manually and local control panel (if design has fan varies from minimum to maximum controlled. local fan control). speed.

viii. Verify the standby RWBVS main Place an AHU in service. Place the MCR display and local, visual supply AHU starts automatically on standby AHU in automatic control. Stop observation indicate the standby AHU the stop of the operating RWBVS the operating recirculation AHU. starts.

main supply AHU.

Tier 2 14.2-59 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-21: Radioactive Waste Building HVAC System Test # 21 (Continued) ix. Verify each standby RWBVS FCU Place an FCU in service. Place the standby MCR display and local, visual starts automatically on the stop of FCU in automatic control. Stop the observation indicate the standby FCU the operating RWBVS fan coil unit. operating FCU. starts.

x. Verify each RWBVS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each RWBVS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific SDIS monitor or an MCR common SDIS monitor if the instrument signal is designed to be displayed on a SDIS monitor.

System Level Test #21-1 Test Objective Test Method Acceptance Criteria

i. Verify the RWB design temperatures Place the RWBVS in automatic operation. i. The temperature and humidity of and humidity monitored by the MCR Place the RXB ventilation system in rooms and areas monitored by the are maintained at design automatic operation. MCR satisfy the design temperature temperature and humidity and humidity requirements conditions during normal operation. contained in Table 9.4.1-2.

ii. Verify the RWBVS maintains a ii. MCR display indicates the RWBVS negative pressure in the RWB relative maintains a negative pressure in the to the outside environment while the RWB relative to the outside RWBVS is operating in normal environment while operating in the alignment. normal operating alignment.

Tier 2 14.2-60 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-22: Turbine Building Ventilation Test # 22 Preoperational test is required to be performed once.

The Turbine Building HVAC system (TBVS) is described in Section 9.4.4 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The TBVS supports the systems located in non-safety related Test #22-1 the Turbine Generator Building (TGB) by providing cooling, heating and humidity control to maintain a suitable environment for the operation of system components.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each TBVS remotely-operated Operate each damper from the MCR and MCR display and local, visual damper can be operated remotely. local control panel (if design has local observation indicate each damper fully damper control). opens and fully closes.

ii. Verify TBVS dampers automatically Open each damper actuated by a smoke MCR display and local, visual close on associated smoke or fire or fire signal. Initiate an alarm signal for observation indicate each damper signals. each damper. closes.

iii. Verify each required TBVS fan stops Initiate an alarm signal for each fan. MCR display and local, visual on actuation of its associated fire or observation indicate each fan stops.

smoke alarm.

iv. Verify the fan speed of each TBVS Vary the speed of each fan from the MCR MCR display indicates the speed of each variable-speed fan can be manually and local control panel (if design has fan varies from minimum to maximum controlled. local fan control). speed.

v. Verify each TBVS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each TBVS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-61 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-22: Turbine Building Ventilation Test # 22 (Continued) vi. Verify each turbine building Initiate a single real or simulated i. The instrument signal is displayed ventilation system (TBVS) instrument instrument signal from each TBVS on an MCR workstation or recorded is monitored in the MCR and the RSS, transmitter. by the applicable control system if the signal is designed to be historian.

displayed in the RSS. ii. The instrument signal is displayed (Test not required if the instrument on an RSS workstation or recorded calibration verified the MCR and RSS by the applicable control system display.) historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #22-1 Test Objective Test Method Acceptance Criteria Verify the Turbine Building battery and Place the turbine bypass system battery The temperature and humidity of battery charger room design and battery charger room ventilation Turbine Building battery and battery temperatures are maintained at design units in automatic operation. charger rooms satisfy the temperature temperature and humidity conditions and humidity requirements.

during normal operation.

Tier 2 14.2-62 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-23: Radioactive Waste Drain System Test # 23 Preoperational test is required to be performed once.

The RWDS is described in Section 9.3.3 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The RWDS supports the RWB by nonsafety-related Test #23-1 collecting radioactive waste in drain sumps and tanks and transfers it to the LRWS for processing.

2. The RWDS supports the RXB by nonsafety-related Test #23-1 collecting radioactive waste in drain sumps and tanks and transfers it to the LRWS for processing.

3. The RWDS supports the ANB by nonsafety-related Test #23-1 collecting radioactive waste in drain sumps and tanks and transfers it to the LRWS for processing.

4. The RWDS supports the UHS by nonsafety-related Test #23-2 providing detection and monitoring of leakage through the UHS liner and the dry dock liner.

5. The LRWS supports the RWDS by nonsafety-related Test #23-1 receiving and processing the effluent from the RWB radioactive waste drain sumps.

6. The LRWS supports the RWDS by nonsafety-related Test #23-1 receiving and processing the effluent from the RXB radioactive waste drain sumps.

7. The LRWS supports the RWDS by nonsafety-related Test #23-1 receiving and processing the effluent from the ANB radioactive waste drain sumps.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each RWDS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each RWDS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each RWDS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each RWDS pump can be Align the RWDS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

v. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from an RWDS grab sample through the grab sampling device. obtained.

device indicated on the RWDS piping and instrumentation diagram.

Tier 2 14.2-63 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-23: Radioactive Waste Drain System Test # 23 (Continued) vi. Verify each RWDS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each RWDS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display). by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #23-1 Test Objective Test Method Acceptance Criteria Verify RWDS pumps start and stop Align each RWDS sump or tank to allow MCR displays and local, visual automatically and transfer liquid waste water in a selected sump or tank to be observation verifies the following:

to its design location. pumped to its design location in the i. The first pump starts on HI level and LRWS (as indicated by the RWDS piping transfers water to its design location and instrumentation diagrams). in the LRWS.

i. Fill the selected sump or tank until a ii. The second (alternate) pump starts HI water level is obtained to start the on HI-HI level.

first (primary) pump. iii. Both primary and alternate pumps ii. Continue filling the sump or tank stop on LO level.

until a HI-HI level starts the second iv. The alternate pump starts on HI (alternate) pump. level.

iii. Stop filling the sump or tank to allow the primary and alternate pumps to stop on low level.

iv. Refill the sump or tank until the alternate pump starts on HI level.

System Level Test #23-2 Test Objective Test Method Acceptance Criteria Verify each RWDS equipment drain sump Fill the selected sump at a rate that PCS data indicates the sump fill rate alarms on a fill rate that exceeds the pool exceeds the PLDS leakage rate setpoint. alarmed at the PLDS leakage rate leakage detection system (PLDS) leakage setpoint.

rate setpoint.

Tier 2 14.2-64 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-24: Balance-of-Plant Drains Test # 24 Preoperational test is required to be performed to support sequence of construction turnover of the BPD system.

BPD system is described in Section 9.3.3 and 11.5.2.2.15 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The BPDS supports the condensate nonsafety-related Test #24-1 polisher demineralizers, the three cooling tower chemical addition systems, and the DWS reverse osmosis units by providing a means to collect and transfer chemical wastes to either the LRWS or to the UWS.

2. The BPDS supports the two TGBs, the nonsafety-related Test #24-1 two diesel generators, the auxiliary boiler, the combustion turbine, the Central Utility Building, and the diesel driven firewater pump by providing a means to collect, treat, and transfer the waste water to the either the LRWS or to the UWS.

3. The BPDS supports the CRB floor nonsafety-related Test #24-1 drains by providing a means to collect, treat, and transfer the waste water to the UWS.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each BPDS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each BPDS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each BPDS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each BPDS pump can be Align the BPDS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

v. Verify the pump speed of each BPDS Vary the speed of each pump from the MCR display indicates the speed of each variable-speed pump can be MCR and local control panel (if design pump varies from minimum to manually controlled. has local pump control). maximum speed.

Tier 2 14.2-65 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-24: Balance-of-Plant Drains Test # 24 (Continued) vi. Verify each BPDS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each BPDS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #24-1 Test Objective Test Method Acceptance Criteria Verify BPDS automatically controlled Align each BPDS sump or tank to allow MCR displays and local, visual pumps start and stop automatically and water in a selected sump or tank to be observation verifies the following:

transfer liquid waste to its design pumped to its design location. If the i. The first pump starts on HI level and location. sump fill rate in the following test transfers water to its design location method is insufficient for automatic start in the LRW system.

of the alternate pump or fire pump, the ii. The second (alternate) pump starts primary pump or alternate pump may be on HI-HI level.

temporarily removed from service to iii. The fire water removal pump starts allow an increase in the sump level.

on HI-HI-HI level (if applicable)

i. Fill the selected sump or tank until a iv. The fire water removal pump stops HI water level is obtained to start the on HI-HI level.

first (primary) pump.

v. Both primary and alternate pumps ii. Continue filling the sump or tank stop on LO level.

until a HI-HI level starts the second (alternate) pump. vi. The alternate pump starts on HI level.

iii. Fill the sump or tank until a HI-HI-HI level starts the fire water removal vii. The primary pump starts on HI-HI pump (if applicable). level.

iv. Stop filling the sump or tank to allow the fire water removal pump to stop on HI-HI level (if applicable).

v. Continue (or start) sump or tank dewatering to allow the primary and alternate pumps to stop on LO level.

vi. Change pump controls to make the alternate pump the first to start, and refill the sump or tank until the first (alternate) pump starts on HI level.

vii. Continue filling the sump or tank until a HI-HI level starts the second (primary) pump.

Tier 2 14.2-66 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-24: Balance-of-Plant Drains Test # 24 (Continued)

System Level Test #24-2 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds Place a north chemical waste water sump i. The north chemical waste water to mitigate a release of radioactivity. pump in operation. Initiate a real or sump pump stops.

simulated high radiation signal on the 6A ii. North chemical waste collection CPS regeneration skid waste effluent. sump to BPDS collection tank isolation valve is closed.

iii. North chemical waste collection sump to LRW high conductivity waste tank isolation valve is closed.

[ITAAC 03.17.02]

(i through iii)

System Level Test #24-3 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds Place a south chemical waste water i. The pump stops.

to mitigate a release of radioactivity. sump pump in operation. Initiate a real or ii. South chemical waste collection simulated high radiation signal on the 6B sump to BPDS collection tank CPS regeneration skid waste effluent. isolation valve is closed.

iii. South chemical waste collection sump to LRW high conductivity waste tank isolation valve is closed.

[ITAAC 03.18.02]

(i through iii)

System Level Test #24-4 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds Place a north waste water sump pump in i. The north waste water sump pump to mitigate a release of radioactivity. operation. Initiate a real or simulated stops.

high radiation signal in the BPDS north ii. North waste water sump discharge TGB floor drains. to BPDS collection tank isolation valve is closed.

iii. North waste water sump discharge to LRW high conductivity waste tank isolation valve is closed.

[ITAAC 03.17.03]

(i thorugh iii)

System Level Test #24-5 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds Place a south waste water sump pump in i. The south waste water sump pump to mitigate a release of radioactivity. operation. Initiate a real or simulated stops.

high radiation signal in the BPDS south ii. South waste water sump discharge TGB floor drains. to BPDS collection tank isolation valve is closed.

iii. South waste water sump discharge to LRW high conductivity waste tank isolation valve is closed.

[ITAAC 03.18.03]

(i through iii)

Tier 2 14.2-67 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-24: Balance-of-Plant Drains Test # 24 (Continued)

System Level Test #24-6 Test Objective Test Method Acceptance Criteria Verify the BPDS automatically responds Place a north waste water sump pump in i. The north chemical waste water to mitigate a release of radioactivity. operation. Initiate a real or simulated sump pump stops.

high radiation signal in the BPDS ii. North chemical waste collection auxiliary blowdown cooler condensate. sump to BPDS collection tank isolation valve is closed.

iii. North chemical waste collection sump to LRW high conductivity waste tank isolation valve is closed.

[ITAAC 03.17.04]

(i through iii)

Tier 2 14.2-68 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-25: Fire Protection System Test # 25 Preoperational test is required to be performed once.

The fire protection system (FPS) is described in Section 9.5.1 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The FPS supports the following nonsafety-related Component-level tests buildings and systems by providing fire prevention, detection, and suppression.

RXB

TGB

RWB

Security Buildings

ANB

Diesel Generator Building

Administration and Training Building

Warehouse Building

Fire Water Building

Switchyard system (SWYD)

Transformers

Site plant cooling structures

Central Utility Building

CRB

MPS

13.8 KV and SWYD system (EHVS)

Medium voltage AC electrical distribution system (EMVS)

Low voltage AC electrical distribution system (ELVS)

RWBVS

CRVS

RBVS

2. The FPS supports the CRB by nonsafety-related Component-level test vii providing audible and visual alarms to alert operators in the MCR.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify an FPS flow balance has been performed.

iii. Verify a pump curve test has been completed for the fire protection pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each FPS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each FPS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each FPS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

Tier 2 14.2-69 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-25: Fire Protection System Test # 25 (Continued) iv. Verify each FPS pump can be started Align the FPS to allow for pump i. MCR display and local, visual and stopped remotely operation. observation indicate each pump i.Start each pump from the MCR and starts.

locally. Audible and visible water hammer ii.Stop each pump locally. are not observed when the pump starts.

ii. MCR display and local, visual observation indicate each pump stops.

v. Verify automatic operation of FPS i. Align the FPS and place the FPS Any MCR display or the local, visual pumps. pumps in automatic operation to observation indicate the following:

pressurize the system. i. The jockey pump maintains the FPS ii. Stop the jockey pump and simulate a header at or greater than 10 psig low FPS header pressure to start the above the pressure setting for the electric fire pump. automatic start of the electric fire iii. Stop the electric fire pump and pump.

simulate a low FPS header pressure ii. The electric fire pump starts.

to start the diesel fire pump. iii. The diesel pump starts.

vi. Verify each valve with a tamper Partially close each FPS manual valve An alarm is received in the MCR when switch alarms when partially closed. with a tamper switch to its alarm position each valve is partially closed.

(approximately 20 per cent of its total travel distance).

vii. Verify each smoke and fire detector Isolate the water supply to each The MCR receives an alarm and provides audible and visual alarms preaction or deluge sprinkler before indication from each smoke and fire and annunciation in the MCR. perming this test to prevent wetting detector.

equipment.

Simulate a smoke or fire signal to each detector.

viii. Verify fire pump flow meets its fire Align the FPS for pump operation i. The electric fire pump meets its protection volumetric flow rate. through the recirculation line. design volumetric flow rate.

i. Start the electric fire pump. [ITAAC 03.07.02]

ii. Start the diesel fire pump.

ii. The diesel fire pump meets its design volumetric flow rate.

[ITAAC 03.07.02]

ix. Verify each FPS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each FPS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-70 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-26: Fire Detection Test # 26 Preoperational test is required to be performed once The fire detection system (FDS) is described in Section 9.5.1 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

As described in Test Abstract Table 14.2- nonsafety-related As described in Test Abstract Table 14.2-25 25 Prerequisites As described in Test Abstract Table 14.2-25 Component Level Tests As described in Test Abstract Table 14.2-25 System Level Tests As described in Test Abstract Table 14.2-25 Tier 2 14.2-71 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-27: Main Steam Test # 27 Preoperational test is required to be performed for each NPM.

The MSS is described in Section 10.3. MS functions are not verified by this test. The MSS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. The MSS supports the SG by nonsafety-related TG Test #33-2 delivering steam to the main condenser.

2. The MSS supports the TGS by nonsafety-related TG Test #33-2 providing steam to the TGS.

Prerequisites Prerequisites associated with MSS testing are identified in the referenced test abstract cited under the Function Verified by Test # heading.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each MSS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

ii. Verify each MSS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each MSS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify automatic operation of MSS Initiate a simulated signal for the Any remote display or local verification extraction steam to protect the main following system conditions. indicates the following:

turbine. i. feedwater heater high level i. extraction steam block valve closes ii. turbine trip ii. extraction steam non-return check valve closes

v. Verify each MSS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each MSS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-72 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-28: Feedwater System Test # 28 Preoperational test is required to be performed for each NPM.

The FWS is described in Section 10.4.7; Section 9.2.6 (condensate storage tank); Section 10.4.1 (condenser); FWS functions are not verified by FWS tests. FWS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. The FWS supports the MSS by nonsafety-related TG Test #33-2 accepting effluent from main steam drains and steam traps into the main condenser.

2. The FWS supports the CPS by nonsafety-related CPS Test #30-1 providing water for CPS rinse and CPS resin transfer.

3. The FWS supports the turbine nonsafety-related CAR Test #32-1 generator by condensing the gland seal steam in the gland exhaust condenser.

4. The FWS supports the CARS by nonsafety-related CAR Test #32-1 accepting water drawn from the exhaust of the vacuum pump.

5. The FWS supports the TG by cooling nonsafety-related TG Test #33-1 superheated steam in the gland steam desuperheater prior to the steam entering the gland seals.

6. The FWS supports the containment nonsafety-related TG Test #33-1 system (CNTS) by supplying feedwater to the SGs.

7. The FWS supports the turbine nonsafety-related TG Test #33-1 generator by cooling superheated turbine bypass steam in the turbine bypass desuperheater prior to the steam entering the main condenser.

8. The FWS supports the turbine nonsafety-related TG Test #33-1 generator by accepting turbine bypass steam into the main condenser.

9. The FWS supports the turbine nonsafety-related TG Test #33-2 generator by accepting exhaust steam from the turbine into the main condenser.

10. The FWS supports the CNTS by nonsafety-related MPS Test #63-5 providing secondary isolation of the feedwater lines.

11. The FWS supports the decay heat nonsafety-related MPS Test #63-5 removal system (DHRS) by providing secondary isolation of the feedwater lines, ensuring required boundary conditions for DHRS operation.

Prerequisites Prerequisites associated with FWS testing are identified in the referenced test abstract cited under the Function Verified by Test # heading.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each FWS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

Tier 2 14.2-73 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-28: Feedwater System Test # 28 (Continued) ii. Verify each FWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each FWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each FWS condensate pump Align the FWS to allow for pump MCR display and local, visual can be started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

v. Verify the pump speed of each FWS Align the FWS to provide a flow path to MCR display indicates the speed of each variable-speed pump can be operate a selected FW variable-speed variable speed pump obtains both manually controlled. pump. minimum and maximum pump speeds.

Vary the FWS pump speed from Audible and visible water hammer are minimum to maximum speed from the not observed when the pump starts.

MCR.

vi. Verify the condensate standby pump Align the FWS to allow for pump MCR display and local, visual automatically starts to protect plant operation. Place a pump in service. observation indicate the standby pump equipment. Initiate a simulated low pump header starts.

pressure low signal. Audible and visible water hammer are not observed when the pump starts.

vii. Verify the feedwater standby pump Align the FWS to allow for pump MCR display and local, visual automatically starts to protect plant operation. Place a pump in service. observation indicate the standby pump equipment. Initiate a simulated low pump header starts.

pressure low signal. Audible and visible water hammer are not observed when the pump starts.

viii. Verify CPS bypass to protect plant Align the FWS to provide a flow path MCR display and local, visual equipment. through the condensate polishers. observation indicate the CPS bypass Initiate a simulated CPS bypass signal. valve is open.

ix. Verify condensate pump low flow i. Align the FWS for automatic short MCR displays and local, visual protection and short cycle automatic cycle cleanup. Place a condensate observation verifies the following:

operation. pump in operation. i. The short cycle flow is automatically ii. Manually throttle a valve in the pump maintained by the short cycle flow path until the flow rate reaches cleanup flow control valve.

the pump minimum flow setpoint. ii. The condensate pump minimum iii. Open the throttled valve. flow valve is open.

iii. The condensate pump minimum flow valve is closed.

x. Verify feedwater pump low flow i. Align the FWS for automatic long MCR displays and local, visual protection. cycle cleanup. Place a condensate observation verifies the following:

pump in operation. i. The long cycle flow is automatically ii. Manually throttle a valve in the pump maintained by the long cycle flow path until the flow rate reaches cleanup flow control valve.

the feedwater pump minimum flow ii. The feedwater pump minimum flow setpoint. valve is open.

iii. Open the throttled valve. iii. The feedwater pump minimum flow valve is closed.

xi. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from an FWS grab sample through the grab sampling device. obtained.

device.

Tier 2 14.2-74 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-28: Feedwater System Test # 28 (Continued) xii. Verify each FWS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each FWS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-75 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-29: Feedwater Treatment Test # 29 Preoperational test is required to be performed for the 6A NPMs and for the 6B NPMs.

The feedwater treatment system (FWTS) is described in Section 10.4.11 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The FWTS supports the FWS by nonsafety-related Component-level tests controlling and maintaining feedwater chemistry within specification.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each FWTS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each FWTS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each FWTS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each FWTS pump can be Align the FWTS to allow for pump MCR display and local, visual started and stopped remotely and operation. observation indicate each pump starts locally (if designed). Stop and start each remotely-controlled and stops.

pump from the MCR.

Stop and start each locally-controlled pump locally.

v. Verify the speed of each FWTS Vary the speed of each pump from the MCR display indicates pump speed variable-speed pump can be MCR and local control panel (if design varies from minimum to maximum manually controlled. has local pump control). speed.

vi. Verify each FWTS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each FWTS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-76 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-30: Condensate Polisher Resin Regeneration System Test # 30 Preoperational test is required to be performed once.

The CPS is described in Section 10.4.6. The CPS and other system functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The CPS supports the FWS by nonsafety-related Test #30-1 providing purification of the feedwater to maintain feedwater chemistry within specification.

2. The FWS supports the CPS by nonsafety-related Test #30-1 providing water for CPS rinse and CPS resin transfer.

3. The ABS supports the CPS by nonsafety-related Test #30-1 supplying steam for resin regeneration.

Prerequisites:

Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each CPS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each CPS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each CPS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each CPS pump can be started Align the CPS to allow for pump MCR display and local, visual and stopped remotely and locally (if operation. observation indicate each pump starts designed). Stop and start each remotely-controlled and stops.

pump from the MCR.

Stop and start each locally-controlled pump locally.

v. Verify the speed of each CPS variable- Vary the speed of each pump from the MCR display indicates pump speed speed pump can be manually MCR and local control panel (if design varies from minimum to maximum controlled. has local pump control) speed.

vi. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a CPS grab sample through the grab sampling device. obtained.

device.

Tier 2 14.2-77 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-30: Condensate Polisher Resin Regeneration System Test # 30 (Continued) vii. Verify each CPS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each CPS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #30-1 Test Objective Test Method Acceptance Criteria Verify the CPS automatically completes Align the FWS to support CPS resin i. The resin transferred to the resin regeneration. regeneration. regeneration skid.

Align the ABS to support CPS resin ii. The CPS regeneration cycle regeneration. completed successfully.

i. Automatically transfer the test resin iii. The resin transferred to a bed from a condensate polisher to condensate polisher.

the CPS regeneration skid.

ii. Initiate an automatic regeneration of the resin.

iii. Automatically transfer the test resin bed from the CPS regeneration skid to a condensate polisher.

Tier 2 14.2-78 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-31: Heater Vents and Drains Test # 31 Preoperational test is required to be performed for each NPM.

The heater vents and drains (HVD) system is described in Section 10.4.7. and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The HVD system supports the FWS by nonsafety-related Component level tests venting the feedwater heaters.

2. The HVD system supports the FWS by nonsafety-related Component level tests controlling level in the shell sides feedwater heaters.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each HVD remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each HVD air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each HVD air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify automatic operation of HVD Initiate a simulated turbine trip. Any remote display or local verification valves to protect the turbine on indicates the following:

turbine trip. i. Low, intermediate and high pressure feedwater heater extraction steam supply valves are closed.

ii. Low, intermediate and high pressure feedwater heater air assisted check valves are closed.

iii. Low, intermediate and high pressure feedwater heater extraction steam dump valves are open.

v. Verify automatic operation of HVD Initiate a simulated signal for the Any remote display or local verification valves to protect turbine on high following system conditions. indicates the following:

feedwater heater level. i. Low pressure feedwater heater high i. Low pressure feedwater heater level. extraction steam supply valve and ii. Intermediate pressure feedwater low pressure feedwater heater heater high level. extraction steam dump valve are iii. High pressure feedwater heater high open.

level ii. Intermediate pressure feedwater heater extraction steam supply valve and intermediate pressure feedwater heater extraction steam dump valve are open.

iii. High pressure feedwater heater extraction steam supply valve and high pressure feedwater heater extraction steam dump valve are open.

Tier 2 14.2-79 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-31: Heater Vents and Drains Test # 31 (Continued) vi. Verify each HVD system instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each HVD system on n MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-80 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-32: Condenser Air Removal System Test # 32 Preoperational test is required to be performed for each NPM.

The condenser air removal system (CARS) is described in Section 10.4.2 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The CARS supports the condensate nonsafety-related Test #32-1 and FWS by removing air and non-condensable gases from the main condenser.

2. The circulating water system (CWS) nonsafety related Test #32-1 supports the FWS by removing heat from the main condenser.

3. The ABS supports the turbine nonsafety-related Test #32-1 generator by supplying gland seal steam.

4. The FWS supports the CARS by nonsafety-related Test #32-1 accepting water drawn from the exhaust of the vacuum pump.

5. The FWS supports the turbine nonsafety-related Test #32-1 generator by condensing the gland seal steam in the gland exhaust condenser.

6. The auxiliary boiler supports the FWS nonsafety-related Test #32-1 by supplying steam to the condenser for sparging when necessary.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each CARS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each CARS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each CARS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each CARS pump can be Align the CARS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

v. Verify CARS valves automatically Initiate a simulated signal for the Any remote display or local verification operate to maintain CARS seal water following system conditions. indicates the following:

separator tank level. i. CARS seal water separator tank high i. CARS seal water separator tank level. makeup valve is closed and drain ii. CARS seal water separator tank low valve is open.

level. ii. The CARS seal water separator makeup valve is open and drain valve is closed.

Tier 2 14.2-81 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-32: Condenser Air Removal System Test # 32 (Continued) vi. Verify a CARS standby pump Align the CARS to allow for pump MCR display and local, visual automatically starts to protect plant operation. Place a pump in service. observation indicate the standby pump equipment. Initiate a simulated main condenser high starts.

pressure. Audible and visible water hammer are not observed when the pump starts.

vii. Verify each CARS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each CARS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #32-1 Test Objective Test Method Acceptance Criteria Verify the CARS can maintain main Place the ABS in automatic control to i. a. The auxiliary boiler gland seal condenser vacuum pressure. supply gland seal steam. steam prevents air leakage into Place the FWS in automatic control to and out of the turbine to condense the gland seal steam in the maintain main condenser design gland exhaust condenser. vacuum pressure.

Place the CWS in automatic control to b. The CWS provides cooling to provide cooling to the main condenser. maintain main condenser design

i. Place the CARS in service to establish vacuum pressure.

vacuum in the main condenser. c. The FWS cools superheated ii. Open the feedwater sparge isolation steam in the gland steam valves to provide steam sparging to desuperheater to design the main condenser. setpoint.

d. The CARS removes noncondensable to maintain main condenser design vacuum pressure.

ii. The ABS is capable of providing sparging steam to the main condenser.

Tier 2 14.2-82 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-33: Turbine Generator Test # 33 Preoperational test is required to be performed for each NPM.

The TGS is described in Sections 10.2, 10.4.3, and 10.4.4. The TGS and other functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The TGS supports the MSS by nonsafety-related Test #33-1 providing steam bypass from the MSS to the main condenser.

2. The MHS supports the CVCS by non-safety related Test #33-1 adding heat to primary coolant.

3. The CVCS supports the reactor nonsafety-related Test #33-1 coolant system (RCS) by heating primary coolant.

4. The ABS supports the module heatup nonsafety-related Test #33-1 system (MHS) by supplying steam for heating reactor coolant at startup and shutdown.

5. The FWS supports the CNTS by nonsafety-related Test #33-1 supplying feedwater to the SGs.

6. The FWS supports the TGS by cooling nonsafety-related Test #33-1 superheated turbine bypass steam in the turbine bypass desuperheater prior to the steam entering the main condenser.

7. The FWS supports the TGS by nonsafety-related Test #33-1 accepting turbine bypass steam into the main condenser.

8. The FWS supports the TGS by cooling nonsafety-related Test #33-1 superheated steam in the gland steam desuperheater prior to the steam entering the gland seals.

9. The FWS supports the TGS by nonsafety-related Test #33-2 accepting exhaust steam from the turbine into the main condenser.

10. The MSS supports the TGS by nonsafety-related Test #33-2 providing steam to the TGS.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

The following prerequisites are not required for component testing:

ii. Verify Test #32-1 has been completed to verify the CARS can maintain main condenser vacuum pressure (reference test 14.2-32).

iii. The SG feedwater flush is complete.

iv. The CARS is automatically maintaining main condenser vacuum.

v. Initial RCS temperature must be approximately 200°F to allow for hot functional testing to obtain data at an RCS temperature of 200°F and above.

vi. The NPM and supporting systems are aligned to increase RCS temperature and pressure.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each TGS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each TGS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

Tier 2 14.2-83 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-33: Turbine Generator Test # 33 (Continued) iii. Verify each TGS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each TGS lube oil pump can be Align the TGS to allow for main lube oil, MCR display and local, visual started and stopped remotely. auxiliary lube oil, and emergency pump observation indicate each pump starts operation. and stops.

Stop and start each pump from the MCR.

v. Verify the TGS exhaust hood is Initiate a simulated high exhaust hood Any remote display or the local, visual protected against high temperature. temperature. observation indicates the exhaust hood spray valve is open.

vi. Verify TGS lubricating oil flow Align the TGS to allow for main lube oil MCR display and local, visual capability by automatic start of the and auxiliary lube oil pump operation. observation indicate the auxiliary oil auxiliary lube oil pump. Place the TGS main oil pump in normal pump starts.

service. Place the auxiliary oil pump in Audible and visible water hammer are standby. not observed when the pump starts.

Simulate a TGS low main oil pump discharge pressure.

vii. Verify TGS lubricating oil flow Align the TGS to allow for auxiliary lube MCR displays and local, visual capability by automatic start of the oil pump and emergency lube oil pump observation indicate the TGS emergency emergency direct current (DC) lube operation. Place the turbine generator oil pump starts.

oil pump. auxiliary oil pump in normal service. Audible and visible water hammer are Simulate a turbine generator auxiliary oil not observed when the pump starts.

pump low discharge pressure or simulate a loss of ac power to start the TGS emergency oil pump.

viii. Verify the turbine stop valve and i. Simulate an overspeed trip signal i. The turbine stop valve and turbine turbine control valves close on from the turbine overspeed control valves close.

turbine overspeed. emergency trip system. [ITAAC 02.04.02]

ii. Simulate an overspeed trip signal from the governor overspeed ii. The turbine stop valve and turbine detection circuit. control valves close.

[ITAAC 02.04.02]

ix. Verify each TGS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each TGS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display). by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-84 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-33: Turbine Generator Test # 33 (Continued)

System Level Test #33-1 Test Objective Test Method Acceptance Criteria

i. Verify the MHS is capable of heating Align the plant to cool the RCS via the i. CVCS supply remains in a sub-the RCS to a temperature sufficient to TGS bypass system. cooled state while heating the RCS obtain criticality. Warm main steam lines. using the module heatup system.

ii. Verify the MHS is capable of heating Place the TGS steam bypass valve in ii. RCS temperature is sufficient to the RCS to establish natural automatic control. obtain criticality.

circulation flow sufficient to obtain Place the feedwater regulating valve in iii. RCS natural circulation flow is criticality. steam generator inventory control. sufficient to obtain criticality.

iii. Verify the TGS automatically controls Place the MHS and the CVCS in automatic iv. The TGS bypass flow is maintained turbine bypass flow to the main control to heat the RCS. at setpoint.

condenser. v. The feedwater flow to the steam Place the ABS high-pressure system in iv. Verify the FWS automatically controls automatic control to heat the MHS heat generator is maintained at setpoint.

flow to the SGs to maintain SG exchanger from RCS ambient vi. The cooled TGS bypass flow is inventory. temperature to the highest temperature maintained at setpoint.

v. Verify the FWS automatically cools achievable by MHS heating. vii. A local grab sample is successfully the TGS bypass steam flow in the obtained at RCS normal operating main steam desuperheater. temperature and pressure.

vi. Verify a local grab sample can be obtained from an MHS system grab sample device.

System Level Test #33-2 This test may be performed after the completion of Test 33-1 when the RCS is at normal operating pressure and the RCS has achieved the maximum temperature achievable by warming the RCS using MHS heating.

Test Objective Test Method Acceptance Criteria Verify the maximum main turbine speed Place the main turbine in service as The maximum main turbine speed is that can be obtained using the MHS to follows: obtained.

heat the RCS. i. Ensure the RCS is at RCS is at normal operating pressure and the RCS is at maximum temperature achievable by warming the RCS using MHS heating.

ii. Place turbine on turning gear with seal steam in service.

iii. Warm up turbine to required temperature.

iv. Increase main turbine speed.

Tier 2 14.2-85 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-34: Turbine Lube Oil System Test # 34 Preoperational test is required to be performed once for the 6A NPMs and once for and the 6B NPMs.

The turbine lube oil storage (TLOS) system is described in Section 10.2.2.1.3 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The TLOS system supports the nonsafety-related Component-level tests turbine generator by supplying lube oil storage for clean and dirty lube oil for protection of plant assets

2. The TLOS system supports the non-safety related Component-level tests turbine generator by receiving and purifying lube oil from the turbine generator lube oil reservoir for protection of plant assets.

3. The TLOS system supports the nonsafety-related Component-level tests turbine generator by providing clean lube oil makeup to the turbine generator lube oil reservoir.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each TLOS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each TLOS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each TLOS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each TLOS pump can be Align the TLOS system to allow for pump MCR display and local, visual started and stopped remotely and operation. observation indicate each pump starts locally (if designed). Stop and start each remotely-controlled and stops.

pump from the MCR.

Stop and start each locally-controlled pump locally.

v. Verify the speed of each TLOS Vary the speed of each pump from the MCR display indicates pump speed variable-speed pump can be MCR and local control panel (if design varies from minimum to maximum manually controlled. has local pump control). speed.

Tier 2 14.2-86 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-34: Turbine Lube Oil System Test # 34 (Continued) vi. Verify each TLOS system instrument Initiate a single real or simulated i. The instrument signal is displayed is monitored in the MCR and the RSS, instrument signal from each TLOS system on an MCR workstation or recorded if the signal is designed to be transmitter. by the applicable control system displayed in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display). by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-87 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-35: Liquid Radioactive Waste System Test # 35 Preoperational test is required to be performed once.

The LRWS is described in Section 11.2 and 11.5.2.1.5 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The LRWS supports the solid nonsafety-related Test #35-1 radioactive waste system (SRWS) by Test #35-2 receiving and processing liquid radioactive waste from the SRWS dewatering skid.

2. The LRWS supports the SFPCS by nonsafety-related Test #35-1 receiving contaminated pool water Test #35-2 to aid in the removal of titrated water or boron. Treated liquid radwaste has the option to return to the pool as makeup.

3. The LRWS supports the CVCS by nonsafety-related Test #35-1 receiving and processing primary Test #35-2 coolant from CVCS letdown.

4. The LRWS supports the RWDS by nonsafety-related Test #35-1 receiving and processing the effluent Test #35-2 from the RWB radioactive waste drain sumps.

5. The LRWS supports the RWDS by nonsafety-related Test #35-1 receiving and processing the effluent Test #35-2 from the RXB radioactive waste drain sumps.

6. The LRWS supports the RWDS by nonsafety-related Test #35-1 receiving and processing the effluent Test #35-2 from the ANB radioactive waste drain sumps.

The LRWS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

The LRWS supports the CVCS by nonsafety-related Test #38-1 receiving and processing primary coolant from CVCS letdown.

The LRWS supports the RWDS by nonsafety-related Test #23-1 receiving and processing the effluent from theRWB radioactive waste drain sumps.

The LRWS supports the RWDS by nonsafety-related Test #23-1 receiving and processing the effluent from theRXB radioactive waste drain sumps.

The LRWS supports the RWDS by nonsafety-related Test #23-1 receiving and processing the effluent from the ANB radioactive waste drain sumps.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each LRWS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control) opens and fully closes.

Tier 2 14.2-88 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-35: Liquid Radioactive Waste System Test # 35 (Continued) ii. Verify each LRWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each LRWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each LRWS pump can be Align the LRWS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

v. Verify the speed of each LRWS Align the LRWS to provide a flow path to MCR display indicates the speed of each variable-speed pump can be operate a selected pump. obtains both minimum and maximum manually controlled. Vary the LRWS pump speed from pump speeds.

minimum to maximum from the MCR. Audible and visible water hammer are not observed when the pump starts.

vi. Verify radiation isolation on Initiate the following a real or simulated MCR display and local, visual discharge to the utility water signals: observation indicate the following:

discharge basin high radiation, low i. LRWS discharge to the utility water i. The LRWS discharge to the utility dilution flow or underground pipe discharge basin high radiation signal. water discharge basin isolation break. ii. LRWS discharge to the utility water valves close.

discharge basin low dilution flow ii. The LRWS discharge to the utility signal. water discharge basin isolation iii. LRWS discharge to the utility water valves close.

discharge basin low guard pipe iii. The LRWS discharge to the utility pressure signal. water discharge basin isolation valves close.

[ITAAC 03.09.07]

(items i through iii) vii. Verify tank valves operate to ensure Simulate an in-service tank high level MCR display and local, visual uninterrupted waste receiving. signal for each of the following tanks: observation indicate the in-service tank low-conductivity waste (LCW) collection fill valve is closed and the standby tank tank A and B fill valve is open.

high-conductivity waste (HCW) collection tank A and B LCW sample tank A and B HCW sample tank A and B viii. Verify degasifier valves operate to i. Initiate a simulated high degasifier i and ii.

ensure uninterrupted waste level signal. MCR display and local, visual receiving. ii. Initiate a simulated high degasifier observation indicate the in-service pressure signal. degasifier fill valve is closed and the standby degasifier fill valve is open.

Tier 2 14.2-89 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-35: Liquid Radioactive Waste System Test # 35 (Continued) ix. Verify LRW pumps automatically Align the LRWS to allow each of the MCR displays and local, visual operate to prevent tank overflow. following LRW transfer pumps to observation indicate the following:

automatically transfer effluent to one of i. The transfer pump starts and its design locations. transfers effluent to its design Degasifier transfer pump A and B location.

LCW collection tank transfer pump A and ii. The transfer pump stops.

B HCW collection tank transfer pump A and B

LCW sample tank transfer pump A and B HCW sample tank transfer pump A and B Detergent waste collection tank transfer pump Demineralized water break tank transfer pump

i. Simulate a HI HI level signal in each of the above tanks.

ii. Simulate a low level signal in each of the above tanks.

x. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a LRWS grab sample through the grab sampling device. obtained.

device indicated on the LRW piping and instrumentation diagram.

xi. Verify SRWS dewatering skid effluent Align SRWS dewatering skid discharge to SRWS dewatering skid effluent is can be transferred to LRW high- one of the LRW high-conductivity waste transferred to the LRW high-conductivity waste (HCW) collection collection tanks. Fill the SRWS conductivity waste collection tank. The tanks. dewatering skid high integrity container SRWS dewatering skid diaphragm pump (HIC) to above the low level pump stop is stopped.

setpoint. Start the SRWS dewatering skid diaphragm pump.

xii. Verify each LRWS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each LRWS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display). by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-90 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-35: Liquid Radioactive Waste System Test # 35 (Continued)

System Level Test #35-1 Test Objective Test Method Acceptance Criteria

i. Verify LRWS can process a gaseous Align LRWS to receive pressurizer i. The LRW degasifier removes waste stream. gaseous waste from the pressurizer condensable gases and vents waste during plant startup. to the RBVS or GRWS.

Process the pressurizer gaseous waste ii. The LRW degasifier liquid transfer through the LRW degasifier. pumps transfer the liquid condensate waste to the low conductivity waste collection tanks.

System Level Test #35-2 Test Objective Test Method Acceptance Criteria

i. Verify LRWS can process a liquid Align LRWS to receive liquid waste from a The waste treatment streams are waste stream. liquid waste stream. successfully processed through the

i. Process the liquid waste stream following processes:

through the low-conductivity waste

filtration (LCW) waste process.

tubular filtration skid ii. Process the liquid waste stream

LCW processing skid through the HCW process.

HCW processing skid

demineralization

transfer to LCW or HCW sample tanks

transfer from LCW or HCW sample tanks to the UWS discharge basin.

Tier 2 14.2-91 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-36: Gaseous Radioactive Waste System Test # 36 Preoperational test is required to be performed once.

The GRWS is described in Section 11.3 and 11.5.2.2.6 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The GRWS supports the LRWS by nonsafety-related Test #36-1 receiving and / or collecting potentially radioactive and hydrogen-bearing waste gases which require processing prior to release to the environment.

2. The GRWS supports the CES by nonsafety-related Test #36-1 receiving and / or collecting potentially radioactive and hydrogen-bearing waste gases which require processing prior to release to the environment.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each GRWS remotely-operated Operate each valve from the (main MCR display and local, visual valve can be operated remotely. control room) MCR and local control observation indicate each valve fully panel (if design has local valve control). opens and fully closes.

ii. Verify each GRWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each GRWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify GRWS valves automatically i. Initiate a real or simulated high GRWS MCR display and local, visual operate to maintain vessel volume. moisture separator level. observation indicate the following:

ii. Initiate a real or simulated low GRWS i. The moisture separator drain valve is moisture separator level. open.

ii. The moisture separator drain valve is closed.

v. Verify GRWS inlet isolation valves Simulate a GRWS inlet stream oxygen MCR display and local, visual automatically close and nitrogen concentration high signal. observation indicate the following:

purge valve opens on high inlet i. The inlet stream isolation valves are stream oxygen concentration. closed.

ii. The nitrogen purge valve is open.

vi. Verify GRWS isolates upon loss of Simulate a loss of RWBVS exhaust flow. MCR display and local, visual RWBV exhaust flow. observation indicate the GRWS isolation valves are closed.

vii. Verify radiation isolation of GRWS i. Initiate a real or simulated GRWS MCR display and local, visual charcoal decay beds upon detection train A decay bed discharge flow observation indicate the following:

of decay bed discharge flow high high radiation signal. i. GRWS train A charcoal decay bed radiation level. ii. Initiate a real or simulated GRWS discharge isolation valve is closed.

train B decay bed discharge flow high [ITAAC 03.09.04]

radiation signal.

ii. GRWS train B charcoal decay bed discharge isolation valve is closed.

[ITAAC 03.09.05]

Tier 2 14.2-92 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-36: Gaseous Radioactive Waste System Test # 36 (Continued) viii. Verify radiation isolation of GRWS Initiate a real or simulated GRWS MCR display and local, visual discharge to the RWBVS exhaust discharge to the RWBVS exhaust high observation indicate the GRWS upon detection of a high radiation radiation signal. discharge to the RWBVS exhaust level. isolation valves are closed.

[ITAAC 03.09.06]

ix. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a GRWS grab sample through the grab sampling device. obtained.

device indicated on the GRWS piping and instrumentation diagram.

x. Verify each GRWS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each GRWS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #36-1 Test Objective Test Method Acceptance Criteria Verify GRWS can process a gaseous waste Align GRWS to receive gaseous waste The gaseous waste stream is successfully stream. from a gaseous waste stream. processed through the following Process the gaseous waste stream processes:

through the gaseous waste process.

gas cooler

moisture separator

charcoal drying heater

charcoal guard bed

charcoal decay beds

RWB exhaust Tier 2 14.2-93 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-37: Solid Radioactive Waste System Test # 37 Preoperational test is required to be performed once.

The SRWS is described in Section 11.4 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The SRWS supports the LRWS by nonsafety-related Test #37-1 receiving spent resin and carbon bed Test #37-4 from LRW processing skids. Test #37-6 Test #37-7

2. The SRWS supports the CVCS by nonsafety-related Test #37-2 receiving spent resin from CVCS ion Test #37-5 exchange vessels. Test #37-7

3. The SRWS supports the PCUS by nonsafety-related Test #37-3 receiving spent resin and sludge Test #37-5 from PCUS ion exchange vessels. Test #37-7

4. The SRWS supports the CRVS by nonsafety-related Test #37-8 receiving exhausted HEPA filters to be compacted and shipped off site.

5. The SRWS supports the RWBVS by nonsafety-related Test #37-8 receiving exhausted HEPA filters to be compacted and shipped off site.

6. The SRWS supports the RBVS by nonsafety-related Test #37-8 receiving exhausted HEPA filters and charcoal bed from RXB and normal control room HVAC, to be compacted and shipped off site.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each SRWS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each SRWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each SRWS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify each SRWS pump can be Align the SRWS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

v. Verify the speed of each SRWS Align the SRWS to provide a flow path to MCR display indicates the speed of each variable-speed pump can be operate a selected pump. obtains both minimum and maximum manually controlled. Vary the SRWS pump speed from pump speeds.

minimum to maximum from the MCR. Audible and visible water hammer are not observed when the pump starts.

vi. Verify each SRWS transfer pump Align the SRWS to allow for transfer MCR display and local, visual automatically stops to protect the pump operation. Place a transfer pump observation indicate each transfer pump pump. in service. Initiate a simulated tank low stopped.

level signal.

Tier 2 14.2-94 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-37: Solid Radioactive Waste System Test # 37 (Continued) vii. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a SRWS grab sample through the grab sampling device. obtained.

device indicated on the SRWS piping and instrumentation diagram.

viii. Verify each SRWS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each SRWS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display). by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #37-1 Test Objective Test Method Acceptance Criteria Verify spent resin from the LRWS Align the LRWS and SRWSs to transfer The waste management control room demineralizers can be transferred to the LRWS demineralizer resin to a SRWS (WMCR) displays and local, visual SRWS phase separator tanks. phase separator tank. observation verifies LRWS demineralizer Start a phase separator transfer pump. resins transferred to a SRWS phase separator tank.

System Level Test #37-2 Test Objective Test Method Acceptance Criteria Verify spent resin from the CVCS ion Align the CVCS and SRWSs to transfer WMCR displays and local, visual exchangers can be transferred to the CVCS ion exchanger resin to a SRWS observation verifies CVCS ion exchanger SRWS spent resin storage tanks. spent resin storage tank. resin transferred to a SRWS spent resin Start a SRWS spent resin storage tank storage tank.

transfer pump.

System Level Test #37-3 Test Objective Test Method Acceptance Criteria Verify spent resin from the PCUS Align the PCUS and SRWSs to transfer WMCR displays and local, visual demineralizers can be transferred to the PCUS demineralizer resin to a SRWS observation verifies PCUS demineralizer SRWS spent resin storage tanks. spent resin storage tank. resins transferred to a SRWS spent resin Start a SRWS spent resin storage tank storage tank.

transfer pump.

System Level Test #37-4 Test Objective Test Method Acceptance Criteria Verify spent resin from the SRWS phase Align a SRW phase separator tank and WMCR displays and local, visual separator tanks can be transferred to a the SRW dewatering station to transfer observation verifies phase separator dewatering station HIC. spent resin to the dewatering station HIC tank resins are transferred to a using service air (SA). dewatering station HIC.

Open SA isolation valve to the SRW phase separator tank.

Tier 2 14.2-95 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-37: Solid Radioactive Waste System Test # 37 (Continued)

System Level Test #37-5 Test Objective Test Method Acceptance Criteria Verify spent resin from the SRW spent Align an SRWS spent resin storage tank WMCR displays and local, visual resin storage tanks can be transferred to and the SRWS dewatering station to observation verifies spent resin storage a dewatering station HIC. transfer spent resin to the dewatering tank resins are transferred to a station HIC using service air. dewatering station HIC.

Open service air isolation valve to the spent resin storage tank.

System Level Test #37-6 Test Objective Test Method Acceptance Criteria Verify granulated activated charcoal from Align a LRWS and SRWSs to granulated WMCR displays and local, visual the LRW granulated activated charcoal activated charcoal to the dewatering observation verifies spent resin storage filter can be transferred to a dewatering station HIC using the clean in place tank resins are transferred to a station HIC. system. dewatering station HIC.

System Level Test #37-7 Test Objective Test Method Acceptance Criteria Verify the dewatering skid pump Align the dewatering skid pump to a Free-standing water in the HIC has been removes standing water in the HIC with LRWS high conductivity waste tank and removed.

spent resin in the dewatering station HIC. start the dewatering skid pump.

System Level Test #37-8 Test Method Acceptance Criteria Verify the SRWS waste compactor Place solid radioactive waste in The waste has been compacted.

compacts solid radioactive waste. compactor and start compactor.

Tier 2 14.2-96 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-38: Chemical and Volume Control System Test # 38 Preoperational test is required to be performed for each NPM.

The CVCS is described in Section 9.3.4 and 11.5.2.2.11 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The CVCS supports the RCS by nonsafety-related Test #38-1 providing primary coolant makeup.

2. The CVCS supports the RCS by nonsafety-related Test #38-1 providing primary coolant letdown.

3. The CVCS supports the RCS by nonsafety-related Test #38-2 providing pressurizer spray flow for RCS pressure control.

4. The CVCS supports the RCS by nonsafety-related Test #38-3 changing the boron concentration of the primary coolant.

5. The BAS supports the CVCS by nonsafety-related Test #38-3 providing uniformly mixed borated water on demand.

6. The LRWS supports the CVCS by nonsafety-related Test #38-1 receiving and processing primary coolant from CVCS letdown.

The CVCS functions verified by other tests are:

The CVCS supports emergency core nonsafety-related MPS Test #63-12 cooling system (ECCS) valves by providing water to reset the ECCS valves.

The CVCS supports the RCS by heating nonsafety-related TG Test #33-1 primary coolant.

The CVCS supports the RCS by isolating safety-related MPS Test #63-6 dilution sources.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify a pump curve test has been completed and approved for the CVCS pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each CVCS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each CVCS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iii. Verify each CVCS ASME Code Class 3 Operate each valve from the MCR. MCR display verifies the valve opens and air-operated valve changes position closes under preoperational under preoperational temperature, temperature, differential pressure, and differential pressure, and flow flow conditions.

conditions. [ITAAC 02.02.03]

iv. Verify each CVCS ASME Code Class 3 Place each valve in its non-safe position. MCR display and local, visual air-operated valve fails to its safe Isolate and vent air to the valve. observation indicate each valve fails to position on loss of air under its safe position under preoperational preoperational temperature, temperature, differential pressure, and differential pressure, and flow flow conditions.

conditions. [ITAAC 02.02.05]

Tier 2 14.2-97 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-38: Chemical and Volume Control System Test # 38 (Continued)

v. Verify each CVCS ASME Code Class 3 Place each valve in its non-safe position. MCR display and local, visual air-operated valve fails to its safe Isolate electrical power to the valve. observation indicate each valve fails to position on loss of electrical power to its safe position under preoperational its solenoid under preoperational temperature, differential pressure, and temperature, differential pressure, flow conditions.

and flow conditions. [ITAAC 02.02.05]

vi. Verify each CVCS ASME Code Class The check valves are tested in Each CVCS ASME Code Class check valve check valve will fully close under accordance with the requirements of strokes fully open and closed under preoperational temperature, ASME OM Code, ISTC-5220, Check Valves. forward and reverse flow conditions, differential pressure and flow respectively.

conditions. [ITAAC 02.02.04]

vii. Verify the speed of each CVCS Align the CVCS to provide a flow path to MCR display indicates the speed of each variable-speed pump can be operate a selected pump. obtains both minimum and maximum manually controlled. Vary the CVCS pump speed from pump speeds.

minimum to maximum from the MCR. Audible and visible water hammer are not observed when the pump starts.

viii. Verify each CVCS operating makeup Align the CVCS to allow for pump MCR display and local, visual pump automatically stops to protect operation. Place a makeup pump in observation indicate the operating the pump and the standby pump service. Initiate a simulated signal for the pump stops and the standby pump starts. following system conditions. starts.

i. Pump low suction pressure. Audible and visible water hammer are ii. Pump high discharge pressure. not observed when the pump starts.

iii. Makeup filter high differential pressure.

ix. Verify each CVCS recirculation pump Align the CVCS to allow for pump MCR display and local, visual automatically stops to protect the operation. Place a recirculation pump in observation indicate the following:

pump and the standby pump starts. service. Initiate a simulated signal for the i. Operating pump stops and the following system conditions. standby pump starts.

i. Pump low suction pressure. ii. Operating pump stops and the ii. Pump discharge high flow. standby pump starts.

iii. High pump discharge pressure. iii. Operating pump stops.

Audible and visible water hammer are not observed when the pump starts.

x. Verify CVCS letdown flow isolates on Initiate a simulated CVCS high letdown MCR display and local, visual high flow to protect plant flow signal. observation indicate the following:

equipment. LRWS letdown flow control valve and LRWS letdown isolation valves (3) are closed.

xi. Verify CVCS hydrogen injection Initiate a simulated CVCS low hydrogen MCR display and local, visual isolates on low injection pressure to injection pressure signal. observation indicate the following:

protect plant equipment. CVCS hydrogen injection pressure regulating valve and hydrogen injection isolation valve are closed.

xii. Verify ion exchanger isolation on Initiate a simulated high non- MCR display and local, visual non-regenerative heat exchanger regenerative heat exchanger outlet observation indicate the following:

high outlet temperature to protect temperature signal. i. CVCS purification bypass diverting plant equipment. valve is in the bypass position.

ii. Mixed bed ion exchanger A inlet isolation valves (2) are closed.

iii. Auxiliary ion exchanger inlet isolation valve is closed.

iv. Cation exchanger inlet isolation valve is closed.

Tier 2 14.2-98 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-38: Chemical and Volume Control System Test # 38 (Continued) xiii. Verify the CVCS automatically Initiate a real or simulated high radiation MCR display verifies the following:

responds to mitigate a release of signal for the auxiliary boiler steam flow i. CVCS module heatup system 6A radioactivity. to the 6A MHS heat exchanger. heat exchanger inlet and outlet isolation valves are closed.

[This component-level test is required to be performed once for each CVCS associated with the MHS 6A heat exchanger.]

[ITAAC 02.07.03]

xiv. Verify the CVCS automatically Initiate a real or simulated high radiation MCR display verifies the following:

responds to mitigate a release of signal for the auxiliary boiler steam flow i. CVCS module heatup system 6B radioactivity. to the 6B MHS heat exchanger. heat exchanger inlet and outlet isolation valves are closed.

[This component-level test is required to be performed once for each CVCS associated with the MHS 6B heat exchanger.]

[ITAAC 02.07.04]

xv. Verify the CVCS automatically Initiate a real or simulated high radiation MCR display verifies the following:

responds to mitigate a release of signal for the RCS discharge flow to the i. chemical and volume control RCS radioactivity. regenerative heat exchanger. discharge to process sampling isolation valve closed.

[This component-level test is required to be performed once for each CVCS.]

[ITAAC 02.07.02]

ix. Verify each CVCS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each CVCS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #38-1 Test Objective Test Method Acceptance Criteria Verify proper operation of the automatic This test will be performed in MCS data indicates that automatic pressurizer level control. conjunction with turbine generator test pressurizer level control is maintained

33-1, which heats the RCS from ambient within the design operating level band.

conditions to no less than 425°F but as high as reasonably achievable.

Use the module control system (MCS) data historian to review pressurizer level at maximum-obtained RCS temperature.

Tier 2 14.2-99 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-38: Chemical and Volume Control System Test # 38 (Continued)

System Level Test #38-2 Test Objective Test Method Acceptance Criteria Verify proper operation of the automatic This test will be performed in MCS data indicates that automatic pressurizer pressure control. conjunction with turbine generator test pressurizer pressure control is

33-1 which heats the RCS from ambient maintained within the design operating conditions to no less than 425°F but as pressure band.

high as reasonably achievable.

Use the MCS data historian to review pressurizer pressure at maximum-obtained RCS temperature.

System Level Test #38-3 Test Objective Test Method Acceptance Criteria Verify proper operation of CVCS This test will be performed in i. MCS data indicates that the dilution automatic dilution and boration control. conjunction with turbine generator test of the RCS results in a decreased

33-1 which heats the RCS from ambient boron concentration within conditions to no less than 425°F but as acceptable limits of the target high as reasonably achievable. concentration.

Ensure that RCS low flow rate alarm is ii. MCS data indicates that the boration clear to ensure adequate mixing for of the RCS results in a increased dilution and boration. boron concentration within

i. Use the MCS automation and acceptable limits of the target operator permission to decrease to a concentration.

target RCS boron concentration.

ii. Use the MCS and operator permission to increase to a target RCS boron concentration.

System Level Test #38-4 Test Objective Test Method Acceptance Criteria Verify the CVCS can provide borated Refer to Test #63-11 Refer to Test #63-11 water to the RCS during a beyond design basis accident.

Tier 2 14.2-100 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-39: Boron Addition System Test # 39 Preoperational test is required to be performed for each NPM.

The boron addition system (BAS) is described in Section 9.3.4. The BAS function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The BAS supports the SFPCS by providing nonsafety-related component-level test xii borated water to the RXB pools.

The BAS function verified by other test is:

System Function System Function Categorization Function Verified by Test #

The BAS supports the CVCS by providing nonsafety-related CVC Test #38-3 uniformly mixed borated water on demand.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify a pump curve test has been completed and approved for the BAS pumps.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each BAS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each BAS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each BAS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iv. Verify the BAS transfer pump can be Align the BAS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate the pump starts Stop and start the transfer pump from and stops.

the MCR. Audible and visible water hammer are not observed when the pump starts.

v. Verify the speed of the BAS variable- Align the BAS to provide a flow path to MCR display indicates the speed of each speed pumps can be manually operate a selected pump. pump obtains both minimum and controlled. Vary the BAS pump speed from maximum pump speeds.

minimum to maximum from the MCR. Audible and visible water hammer are not observed when the pump starts.

vi. Verify BAS valves automatically i. Initiate a real or simulated high BAS MCR display and local, visual operate to protect plant equipment. batch tank level signal. observation indicate the following:

ii. Initiate a real or simulated high BAS i. The batch tank fill and return valves storage tank level signal. are fully closed.

ii. The storage tank fill and recirculation valves are fully closed.

vii. Verify the BAS transfer pump stops Align the BAS to allow for pump MCR display and local, visual automatically to protect plant operation. observation indicate the following:

equipment. i. Place the BAS transfer pump in i. The transfer pump stops.

service on recirculation to the BAS ii. The transfer pump stops.

batch tank. Initiate a simulated a low batch tank level signal.

ii. Place the BAS transfer pump in service on recirculation to the BAS storage tank. Simulate a low storage tank level signal.

Tier 2 14.2-101 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-39: Boron Addition System Test # 39 (Continued) viii. Verify BAS supply pumps stop Align the BAS to allow for pump MCR display and local, visual automatically to protect plant operation. observation indicate the following:

equipment. i. Place a BAS supply pump in service i. The supply pump stops.

on recirculation to the BAS batch ii. The supply pump stops.

bank. Initiate a simulated a low batch tank level signal.

ii. Place a supply pump in service on recirculation to the BAS storage tank.

Initiate a simulated low storage tank level signal.

ix. Verify BAS flow capability by Align the BAS to allow for pump MCR display and local, visual automatic start of each BAS supply operation. Place a supply pump in observation indicate the standby pump pump while in standby mode. service. Initiate a simulated low pump starts.

discharge pressure signal. Audible and visible water hammer are not observed when the pump starts.

x. Verify supply pump low flow Align the BAS to allow a BAS supply MCR displays and local, visual protection. pump flow sufficient to close the pump observation verifies the following:

recirculation valve to the storage tank. i. The pump recirculation valve is

i. Manually throttle a valve in the pump open.

flow path until the flow rate reaches ii. The pump recirculation valve is the pump minimum flow setpoint. closed.

ii. Open the throttled valve xi. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a BAS grab sample through the grab sampling device. obtained.

device.

xii. Verify the BAS automatically adds a i. Verify the BAS batch tank contains a MCR displays and local, visual specified quantity of borated water sufficient volume of water to conduct observation verifies the following:

from the BAS batch tank to the RXB this test. i. The BAS to SFPCS valve initially pools. ii. Align the BAS and the SFPCS to opens to supply water from the BAS supply water from the BAS to the to the SFPCS pump suction.

SFPCS pump suction. ii. The BAS to SFPCS valve iii. Enter a BAS batch tank target level to automatically closes when the BAS terminate batch operation to the batch tank obtains the target level.

spent fuel pool.

xiii. Verify each BAS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each BAS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-102 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-40: Module Heatup System Test # 40 Preoperational test is required to be performed for each NPM.

The MHS is described in Section 9.3.4.2. MHS functions are not verified by MHS tests. MHS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

The MHS supports the CVCS by adding nonsafety-related TG Test #33-1 heat to primary coolant.

Prerequisites Prerequisites associated with MHS testing are identified in the referenced test abstract cited under the Function Verified by Test # heading.

Component Level Tests None System Level Tests None Tier 2 14.2-103 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-41: Containment Evacuation System Test # 41 Preoperational test is required to be performed for each NPM.

The CES is described in Sections 9.3.6, 11.5.2.2.7 and 5.2.5 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The CES supports the CNTS by nonsafety-related Test #41-1 removing water vapor from the Test #41-2 containment vessel (CNV). Test #41-3

2. The CES supports the CNTS by nonsafety-related Test #41-1 condensing water vapor removed Test #41-2 from the CNV in the containment Test #41-3 evacuation condenser.

3. The CES supports the CNTS by nonsafety-related Test #41-1 removing non-condensable gases Test #41-2 from the CNV. Test #41-3

4. The CES supports the MCS by nonsafety-related Test #41-2 providing a radioactivity signal.

5. The CES supports the RCS by nonsafety-related Test #41-3 providing RCS leak detection monitoring capability.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each CES remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each CES air-operated valve Place each CES valve in its non-safe MCR display and local, visual fails to its safe position on loss of air. position. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each CES air-operated valve Place each CES valve in its non-safe MCR display and local, visual fails to its safe position on loss of position. Isolate electrical power to each observation indicate each valve fails to electrical power to its solenoid. CES air-operated valve. its safe position.

iv. Verify each CES pump can be started Stop and start each pump from the MCR. MCR display and local, visual and stopped remotely. observation indicate each pump starts and stops.

v. Verify the speed of each CES variable- Vary the speed of each pump from the MCR display indicates pump speed speed pump can be manually MCR and local control panel (if design varies from minimum to maximum controlled. has local pump control). speed.

vi. Verify each CES pump automatically Place a pump in operation. Initiate a real MCR displays and local, visual stops to protect plant equipment. or simulated signal for each pump trip observation verifies the pump stops.

condition.

vii. Verify each CES pump suction and Open the pump suction and discharge Each pump suction and discharge valve discharge valve automatically closes valves. closes on each real or simulated valve to protect the CES equipment. Initiate a real or simulated signal for each close condition.

valve close conditions.

viii. Verify a local grab sample can be Place the system in service to allow flow A local grab sample is successfully obtained from a CES grab sample through the grab sampling device. obtained.

device indicated on the CES piping and instrumentation diagram.

Tier 2 14.2-104 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-41: Containment Evacuation System Test # 41 (Continued) ix. Verify each CES instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each CES on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #41-1 Test Objective Test Method Acceptance Criteria Verify the automatic operation of the CES After the containment flooding and drain The automated control establishes and to establish and maintain design vacuum system (CFDS) completes draindown of maintains vacuum in the CNV.

for the CNV. the CNV, place the CES in automatic operation.

System Level Test #41-2 Test Objective Test Method Acceptance Criteria Verify radiation isolation and flow The NPM is in hot functional testing with i. The CES effluent flow path to the diversion on high radiation level in the the RCS at normal operating pressure. RBVS is isolated and diverted to CES. The CES is operating in automatic control GRWS.

with a CNV steady-state vacuum pressure [ITAAC 02.07.01]

indicating the noncondensable gases have been removed from the CNV.

ii. The CES effluent to process sample Initiate a real or simulated high radiation panel isolation valve is closed.

signal for the CES vacuum pump

[ITAAC 02.07.01]

discharge.

iii. The CES purge air solenoid valves to the vacuum pumps are closed.

[ITAAC 02.07.01]

Tier 2 14.2-105 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-41: Containment Evacuation System Test # 41 (Continued)

System Level Test #41-3 Test Objective Test Method Acceptance Criteria

i. Verify the CES level instrumentation i. The NPM is in hot functional testing i. The CES detects a level increase in supports RCS leakage detection. with the RCS at normal operating the CES sample tank, which ii. Verify the CES pressure pressure and the maximum correlates to a detection of an instrumentation supports RCS operating temperature achievable by unidentified RCS leakage rate of one leakage detection. heating the RCS with the MHS. gpm within one hour, by providing ii. The CES is operating in automatic an alarm signal to the MCR within control with a CNV steady-state one hour of the start of water vacuum pressure indicating the injection into the CNV indicating the noncondensable gasses have been baseline leakage rate has been removed from the CNV. exceeded.

iii. Record the MCS baseline leakage rate [ITAAC 02.03.01]

into the CNV. ii. The CES detects a pressure increase iv. Isolate the CFDS to CNTS spool piece in the sample vessel inlet pressure to allow test equipment to be instrumentation that correlates to a connected to the spool piece. detection of an unidentified RCS

v. Inject water at a flow rate less than or leakage rate of one gpm within one equal to one gpm. hour, by providing an alarm signal to the MCR within one hour of the start This test may be done in conjunction of water injection into the CNV Test #41-2.

indicating the baseline leakage rate has been exceeded.

[ITAAC 02.03.02]

Tier 2 14.2-106 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-42: Containment Flooding and Drain System System Test # 42 Preoperational test component level testing is required to be performed for the 6A CFDS and for the 6B CFDS. System level testing is required to be performed as indicated for each system level test.

The CFDS is described in Section 9.3.6 and 11.5.2.2.9 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The CFDS supports the CNTS by nonsafety-related Test #42-1 flooding the CNV in preparation for refueling operations.

2. The CFDS supports the CNTS by nonsafety-related Test #42-2 draining the CNV in preparation for startup operations.

3. The CFDS supports the RCS by nonsafety-related Test #42-3 providing borated coolant inventory for the removal of core heat during a beyond design basis accident.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each CFDS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each CFDS air-operated valve Place each CFDS valve in its non-safe MCR display and local, visual fails to its safe position on loss of air. position. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each CFDS air-operated valve Place each CFDS valve in its non-safe MCR display and local, visual fails to its safe position on loss of position. Isolate electrical power to each observation indicate each valve fails to electrical power to its solenoid. CFDS air-operated valve. its safe position.

iv. Verify each CFDS pump can be Stop and start each pump from the MCR. MCR display and local, visual started and stopped remotely. observation indicate each pump starts and stops.

v. Verify each CFDS pump Place a pump in operation. Initiate a real MCR displays and local, visual automatically stops to protect plant or simulated signal for each pump trip observation verifies the pump stops.

equipment. condition.

vi. Verify each CFDS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each CFDS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-107 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-42: Containment Flooding and Drain System System Test # 42 (Continued)

System Level Test #42-1 Test Objective Test Method Acceptance Criteria Verify the CFDS can automatically drain Drain the CNTS using CFDS automatic The CNTS is drained using CFDS the CNTS. operation and designed manual automatic controls.

operation. (This test is required to be performed for each NPM.)

System Level Test #42-2 Test Objective Test Method Acceptance Criteria Verify the CFDS can automatically flood Drain the CNTS using CFDS automatic The CNTS is flooded using CFDS the CNTS. operation and designed manual automatic controls.

operation. (This test is required to be performed for each NPM.)

System Level Test #42-3 Test Objective Test Method Acceptance Criteria Verify the CFDS can provide borated Refer to Test #63-11 Refer to Test #63-11 water to the containment during a beyond design basis accident.

System Level Test #42-4 Test Objective Test Method Acceptance Criteria Verify the 6A CFDS automatically While the 6A CFDS is draining the CNTS The 6A CFDS containment drain responds to mitigate a release of initiate a real or simulated high radiation separator gaseous discharge to RBVS radioactivity. signal on the gaseous effluent of the 6A isolation valve is closed.

CFDS containment drain separator tank. [ITAAC 03.17.01]

System Level Test #42-5 Test Objective Test Method Acceptance Criteria Verify the 6B CFDS automatically While the 6B CFDS is draining the CNTS The 6B CFDS containment drain responds to mitigate a release of initiate a real or simulated high radiation separator gaseous discharge to RBVS radioactivity. signal on the gaseous effluent of the 6B isolation valve is closed.

CFDS containment drain separator tank. [ITAAC 03.18.01]

Tier 2 14.2-108 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-43: Containment System Test # 43 Preoperational test is required to be performed for each NPM.

The CNTS is described in Section 6.2 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The CNTS supports the RXB by safety-related Test #43-1 providing a barrier to contain mass, energy, and fission product release from a degradation of the reactor coolant pressure boundary.

2. The CNTS supports the ECCS safety-related Test #43-1 operations by providing a sealed containment.

The CNTS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

The CNTS supports the DHRS by closing safety-related MPS Test #63-6 containment isolation valves (CIVs) for the main steam and feedwater systems when actuated by the MPS for DHRS operation.

The CNTS supports the RCS by closing safety-related MPS Test #63-6 the CIVs for pressurizer spray, RCS injection, RCS discharge, and reactor pressure vessel (RPV) high point degasification when actuated by the MPS for RCS isolation.

The CNTS supports the RXB by providing safety-related MPS Test #63-6 a barrier to contain mass, energy, and fission product release by closure of the CIVs upon a containment isolation signal.

The CNTS supports the Reactor Building non-safety related, risk-significant RBC Test #52-1 crane (RBC) by providing lifting RBC Test #52-2 attachment points that the RBC can connect to so that the module can be lifted.

The CNTS supports the MPS by providing non-safety related SDI Test #66-2 post-accident monitoring (PAM) nonsafety-related information signals Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify the CNTS safety-related check The check valves are tested in Each CNTS safety-related check valve valves change position under design accordance with the requirements of strokes fully open and closed under differential pressure and flow. ASME OM code, ISTC-5220, check valves. forward and reverse flow conditions, respectively.

[ITAAC 02.01.21]

System Level Test #43-1 Test Objective Test Method Acceptance Criteria Verify the leaktightness of the Perform 10 CFR Part 50, Appendix J local Local leak rate tests are completed on containment system. leak rate tests (Type B and Type C tests) containment penetrations listed in Table of the CNTS in accordance with the 6.2-9 which require Appendix J, Type B guidance provided in ANSI/ANS 56.8,RG or C testing.

1.163, and NEI 94-01.

[ITAAC 02.01.07]

Tier 2 14.2-109 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-44: Control Rod Drive System Flow-Induced Vibration Test # 44 Validation testing is performed during factory testing on the control rod drive shaft per Table 4-1 of TR-0716-50439. There are no preoperational tests for CRDS.

The CRDS flow-induced vibration testing is performed consistent with the requirements of the NuScale Comprehensive Vibration Assessment Program as described in the Comprehensive Vibration Assessment Program (CVAP) Technical Report, TR-0716-50439. Visual examination of the CRDS components is performed as specified in Table 5-1 of TR-0716-50439. This test is coordinated with Test #108. The CVAP is addressed in Section 3.9.2. The CRDS is discussed in Section 4.6.

System Function System Function Categorization Function Verified by Test #

None N/A N/A Prerequisites:

N/A Component Level Tests None Tier 2 14.2-110 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-45: Reactor Vessel Internals Flow-Induced Vibration Test # 45 Validation testing is performed at the factory for the in-core instrument guide tubes per Table 4-1 of TR-0716-50439. There are no preoperational tests for RVI.

RVI flow-induced vibration testing is performed consistent with the requirements of the NuScale Comprehensive Vibration Assessment Program as described in the Comprehensive Vibration Assessment Program (CVAP) Technical Report, TR-0716-50439. Visual examination of the RVI components is performed as specified in Table 5-1 of TR-0716-50439. This test is coordinated with Test #108. The CVAP is addressed in Section 3.9.2. Reactor vessel internals are discussed in Section 5.1.3.3.

System Function System Function Categorization Function Verified by Test #

None N/A N/A Prerequisites:

N/A Component Level Tests None Tier 2 14.2-111 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-46: Reactor Coolant System Test # 46 Preoperational test is required to be performed for each NPM.

The RCS is described in Section 5.4 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

The RCS supports the CNTS by supplying safety-related, risk-significant Component Tests i. and ii.

the reactor coolant pressure boundary and a fission product boundary via the RPV and other appurtenances.

The RCS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

The RCS supports the MPS by providing safety-related, risk-significant MPS Test #63-1 instrument information signals for MPS actuation.

The RCS supports the MPS by providing safety-related MPS Test #63-1 instrument information signals for low temperature overpressure protection actuation.

The RCS supports the MPS by providing nonsafety-related SDI Test #66-2 PAM instrument information signals.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify the RCS safety-related check The check valves are tested in Each RCS safety-related check valve valves change position under design accordance with the requirements of strokes fully open and closed under differential pressure and flow. ASME OM Code, ISTC-5220, Check Valves. forward and reverse flow conditions, respectively.

[ITAAC 02.01.16]

ii. Verify the RCS safety-related excess The check valves are tested in Each RCS safety-related excess flow flow check valves change position accordance with the requirements of check valve strokes fully closed under under design flow. ASME OM Code, ISTC-5220, Check Valves. excess flow conditions.

[ITAAC 02.01.17]

iii. Verify each RCS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each RCS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to b displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-112 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-47: Emergency Core Cooling System Test # 47 Preoperational test is required to be performed for each NPM.

The ECCS is described in Section 6.3. The ECCS functions are not verified by ECCS tests. The ECCS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. The ECCS supports the RCS by safety-related MPS Test #63-6 opening the ECCS reactor vent valves and reactor recirculation valves when their respective trip valve is actuated by the MPS.

2. The ECCS supports the RCS by safety-related MPS Test #63-6 providing recirculated coolant from the containment to the RPV for the removal of core heat.

3. The ECCS supports the RCS by safety-related MPS Test #63-6 providing low temperature overpressure protection (LTOP) for maintaining the reactor coolant pressure boundary.

4. ECCS supports MPS by providing nonsafety-related MPS Test #63-1 instrumentation information signals.

5. The ECCS supports MPS by providing nonsafety related SDI Test #66-2 post accident monitoring instrument information signals.

Prerequisites Prerequisites associated with ECCS testing are identified in the referenced test abstract cited under the Function Verified by Test # heading.

Component Level Tests None System Level Tests None Tier 2 14.2-113 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-48: Decay Heat Removal System Test # 48 Preoperational test is required to be performed for each NPM.

The DHRS is described in Section 6.3. DHRS functions are not verified by DHRS tests. DHRS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. The DHRS supports the RCS by safety-related MPS Test #63-6 opening the DHRS actuation valves for DHRS operation.

2. The DHRS supports the MPS by safety-related MPS Test #63-1 providing MPS actuation instrument information signals.

3. The DHRS supports the MPS by nonsafety-related SDI Test #66-2 providing PAM instrument information signals.

Prerequisites N/A Component Level Tests None System Level Tests None Tier 2 14.2-114 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-49: Incore Instrumentation Test # 49 Preoperational test is required to be performed for each NPM.

The in-core instrumentation system (ICIS) is described in Section 7.0.4.7 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The IICIS supports the MPS by providing nonsafety-related Test #49-1 reactor core (RXC) temperature information.

The ICIS functions verified by another test is:

System Function System Function Categorization Function Verified by Test #

The ICIS supports the MPS by providing nonsafety-related Test #66-2 RXC temperature information.

Prerequisites

i. The ICIS instrument strings are inserted into the core.

i. Verify an instrument calibration has been performed on all ICIS thermocouples by cross-calibrating the thermocouple to the RCS narrow range resistance temperature detectors (RTDs) prior to RCS heatup.

Component Level Tests None System Level Test #49-1 Test Objective Test Method Acceptance Criteria Verify proper temperature indication is Heat the RCS from ambient conditions to MCS data indicates that the ICIS obtained from the ICIS thermocouples. normal operating temperature. thermocouples respond properly.

Use the MCS data historian to cross-check the ICIS thermocouples to each other and the RCS narrow-range and wide range RTDs.

Tier 2 14.2-115 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-50: Module Assembly Equipment Test # 50 There are no preoperational tests for module assembly equipment (MAE).

The MAE consists of module import trolley, the upender, and the inspection rack.

System Function System Function Categorization Function Verified by Test #

None None None Prerequisites N/A Component Level Tests None System-Level Tests None Tier 2 14.2-116 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-51: Fuel Handling Equipment System Test # 51 Component-level testing is required to be performed once.

System Level Test #51-1 and Test #51-2 are required to be performed once.

System Level Test #51-3 and Test #51-4 are required to be performed once.

The fuel handling equipment (FHE) system is described in Section 9.1.4 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The FHE system supports new fuel by nonsafety-related Test #51-1 providing ability to visually inspect Test #51-2 fuel.

2. The FHE system supports the RXC by nonsafety-related Test #51-3 moving fuel within the core. Test #51-4

3. The FHE system supports the spent nonsafety-related Test #51-4 fuel storage system (SFSS) by moving fuel into the spent fuel storage system.

Prerequisites

i. An FHE system factory acceptance test has been successfully completed and approved.

ii. A rated-load test has been successfully completed and approved on theFHE system on the following equipment in accordance with ASME NOG-1 paragraph 7423.

a. Fuel handling machine (FHM) main hoist

b. FHM auxiliary hoists iii. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify the operation of FHE controls Actuate or simulate actuation of the The FHE equipment controls limit that limit motion and speed. interlocks contained in Table 14.2-51a. motion and speed per design.

ii. Verify each FHE system instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each FHE system on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #51-1 Test Objective Test Method Acceptance Criteria Verify the proper operation of the new Transfer a dummy fuel assembly from its i. A dummy fuel assembly is fuel jib crane. receipt shipping container to the new successfully transferred to the new fuel inspection stand and from the new fuel inspection stand.

fuel inspection stand to the new fuel ii. A dummy fuel assembly is elevator. successfully transferred to the new fuel elevator.

Tier 2 14.2-117 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-51: Fuel Handling Equipment System Test # 51 (Continued)

System Level Test #51-2 Test Objective Test Method Acceptance Criteria Verify the proper operation of the new Lower a dummy fuel assembly in the new A dummy fuel assembly is successfully fuel elevator. fuel elevator. lowered to the position where it can be retrieved by the FHM mast.

System Level Test #51-3 Test Objective Test Method Acceptance Criteria Verify the proper operation of the FHM. i. Transfer the dummy fuel assembly i. The dummy fuel assembly is from the new fuel elevator to the successfully transferred to the FHM FHM mast. mast.

ii. Transfer the dummy fuel assembly ii. The dummy fuel assembly is from the new fuel elevator location successfully transferred to its to a designated RXC location. designated core location and iii. Seat the dummy fuel assembly. partially inserted.

iii. The dummy fuel assembly is fully seated.

System Level Test #51-4 Test Objective Test Method Acceptance Criteria Verify the proper operation of the FHM. i. Withdraw the dummy fuel assembly i. The dummy fuel assembly is to a position where the FHM can successfully transferred to its automatically transfer the assembly. designated storage location and Transfer the dummy fuel assembly partially inserted.

from the RXC to a designated spent ii. The dummy fuel assembly is fully fuel storage location. seated.

(Manual operation of the fuel assembly is required for final fuel insertion.)

ii. Seat the dummy fuel assembly.

System Level Test #51-5 Test Objective Test Method Acceptance Criteria Verify the FHM maintains at least 10 feet Perform a test of the FHM mast The FHM maintains at least 10 feet of of water above the top of the fuel mechanical stop limit switch. water above the top of the fuel assembly when lifted to its maximum assembly when lifted to its maximum height with the pool level at the lower height with the pool level at the lower limit of the normal operating low water limit of the normal operating low water level. level.

[ITAAC 03.04.05]

System Level Test #51-6 Test Objective Test Method Acceptance Criteria The new fuel jib crane hook movement is Using the new fuel jib crane hook The new fuel jib crane interlocks prevent limited to prevent carrying a fuel attempt to transfer a dummy fuel the crane from carrying a fuel assembly assembly over the fuel storage racks in assembly or new fuel assembly over the over the spent fuel racks.

the spent fuel pool. fuel storage racks in the spent fuel pool. [ITAAC 03.04.06]

Tier 2 14.2-118 Revision 1

Tier 2 NuScale Final Safety Analysis Report Table 14.2-51a: FHE System Interlock Testing Equipment Emergency Bridge and Crane Zone Hoist Underload Hoist Hoist Up- Hoist Down- Rotation Hoist, Overspeed Mis-reeve Stop Trolley End Limits (Underweight / Overload Position Position and Bridge, Limit Limit Switch of Travel Slack Rope) (Overweight) (Upper (Lower Gripper Trolley (Lower) (Raise) Travel Travel Slow Zones Limit) Limit)

FHM bridge X X X --- --- --- --- X --- ---

FHM trolley X X X --- --- --- --- X --- ---

FHM main X --- --- X X X X X X X X hoist/mast FHM X --- --- --- X X X --- --- X auxiliary hoist New fuel jib X X --- --- --- --- --- X --- ---

crane trolley New fuel jib X --- --- --- X X X X X crane hoist New fuel X --- --- --- X X X X 14.2-119 elevator Initial Plant Test Program Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-52: Reactor Building Cranes Test # 52 Preoperational test is required to be performed once.

The RBC system is described in Section 9.1.5 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The RBC supports the NPM by nonsafety-related, risk-significant Test #52-1 providing structural support and Test #52-2 mobility while moving from refueling, inspection and operating bay.

2. MAE bolting supports the CNT by nonsafety-related Test #52-2 providing material handling to allow for disassembly and reassembly of the CNV lower flange.

3. MAE bolting supports the RPV nonsafety-related Test #52-2 actively by providing material handling to allow for disassembly and reassembly of the RPV lower flange.

4. The CNTS supports the RBC by nonsafety-related Test #52-1 providing lifting attachment points Test #52-2 that the RBC can connect to so that the module can be lifted.

Prerequisites

i. An RBC site acceptance test has been completed and approved.

ii. A rated-load test has been completed and approved on the RBC on the following equipment in accordance with ASME NOG-1 paragraph 7423.

a. RBC main hoist

b. RBC auxiliary hoists

c. RBC wet hoist iii. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify RBC controls that limit RBC Actuate or simulate actuation of the RBC Local visual observation indicates that motion and speed. interlocks contained in Table 14.2-52a. the interlocks limit RBC motion and speed.

ii. Verify RBC remains in current Initiate the following real or simulated Local visual observation indicates that position on loss of control or power signals: the bridge, trolley, main hoist, wet hoist, or seismic event. i. Loss of control. auxiliary hoist trolley and auxiliary hoist ii. Loss of power. brakes are set.

iii. Seismic switch actuation.

iii. Verify each RBC system instrument is Initiate a single real or simulated i.The instrument signal is displayed on monitored in the MCR, if designed to instrument signal from each RBC system an MCR workstation or recorded by the be monitored in the MCR. transmitter. applicable control system historian.

(Test not required if the instrument calibration verified the MCR display)

Tier 2 14.2-120 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-52: Reactor Building Cranes Test # 52 (Continued)

System Level Test #52-1 Test Objective Test Method Acceptance Criteria

i. Verify RBC load path and removal of Place the module lifting adaptor on the i. The bridge and trolley speeds do not an NPM from a reactor bay. RBC. Lift an NPM and move the RBC with exceed maximum design speeds.

ii. Verify RBC load path and installation the attached NPM to its design home ii. The bridge and trolley does not of an NPM in a reactor bay. location. move at the same time.

i. Use the RBC semi-automatic iii. The bridge and trolley maximum programmed controls to install the allowable speed is toggled from full-NPM in the lead NPM bay location speed to microspeed when the RBC and return the RBC to the design hook gets within the design home location distance of a predefined reference ii. Use the RBC semi-automatic location.

programmed controls to retrieve the iv. The main hoist only moves within NPM from the lead NPM bay location the predefined elevation zones.

and return the RBC with attached v. The NPM is positioned at the design module to the design home location. rotation at predefined reference Repeat this sequence relative to the locations.

installation of additional NPMs. vi. The NPM is fully seated in the reactor bay receiver.

i. a. Verify the NPM can be The RBC is at the design home location i. a. Verify the NPM can be disassembled using the CNV with an NPM attached to the module disassembled using the CNV support stand and the RPV lifting adaptor (MLA). support stand and the RPV support stand and associated i. Use the RBC semi-automatic support stand and associated tooling. programmed controls to move the tooling.

b. Verify the RBC semi-automatic NPM from the design home location b. Verify the RBC semi-automatic controls can be used to transport to the CNV support stand and seat controls can be used to the NPM through the the NPM lower CNV in the CNV transport the NPM through the disassembly process. support stand. De-tension and disassembly process.

ii. a. Verify the NPM can be assembled remove the lower CNV closure bolts. ii. a. Verify the NPM can be using the CNV support stand and Use the RBC semi-automatic assembled using the CNV the RPV support stand and programmed controls to move the support stand and the RPV associated tooling. NPM from the CNV support stand to support stand and associated

b. Verify the RBC semi-automatic the RPV support stand and seat the tooling.

controls can be used to transport NPM in the RPV support stand. De- b.Verify the RBC semi-automatic the NPM through the assembly tension and remove the lower RPV controls can be used to transport process. closure bolts. the NPM through the assembly Use the RBC semi-automatic process.

programmed controls to move the upper NPM from the RPV support stand to the module inspection rack and seat the upper NPM on the module inspection rack support lug receiving pockets.

Use the RBC semi-automatic programmed controls to disengage the MLA from the upper NPM and move the RBC and MLA from the module inspection rack to the design home location.

Tier 2 14.2-121 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-52: Reactor Building Cranes Test # 52 (Continued)

System Level Test #52-2 Test Objective Test Method Acceptance Criteria ii. Use the RBC semi-automatic programmed controls to move the NPM and MLA from the design home location to the module inspection rack and attach the upper NPM to the MLA.

Use the RBC semi-automatic programmed controls to move the upper NPM from the module inspection rack to the RPV support stand and seat the upper NPM on the lower RPV and RPV support stand.

Install and tension the lower RPV closure bolts.

Use the RBC semi-automatic programmed controls to move the upper NPM from the RPV support stand to the CNV support stand and seat the upper NPM on the lower CNV and CNV support stand. Install and tension the lower CNV closure bolts.

Use the RBC semi-automatic programmed controls move the RBC and NPM from the CNV support stand to the design home location.

Tier 2 14.2-122 Revision 1

Tier 2 NuScale Final Safety Analysis Report Table 14.2-52a: RBC System Interlock Testing Equipment Emergency Bridge and Crane Zone Hoist Underload Hoist Hoist Up- Hoist Down- Overspeed Mis-reeve Unbalanced Two-Stop Trolley End Limits (Underweight / Overload Position Position Limit Limit Switch Load Blocking of Travel Slack Rope) (Overweight) (Upper (Lower (Lower) (Raise) Travel Travel Limit) Limit)

RBC trolley X X X RBC bridge X X X RBC main X X X X X X X X X hoist RBC aux X X hoist trolley 1&2 RBC aux X X X X X X X X hoist 1 & 2 Wet hoist X X X X X X X X X 14.2-123 Initial Plant Test Program Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-53: Process Sampling System Test # 53 Preoperational test is required to be performed for each NPM.

The PSS is described in Section 9.3.2 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The PSS supports the RCS during nonsafety-related Test #53-1 normal operations by providing sampling and analysis of reactor coolant discharge (letdown) liquid.

2. The PSS supports the CVCS by nonsafety-related Test #53-1 providing sampling of reactor coolant at process points in the CVCS.

3. The PSS supports the RCS during nonsafety-related Test #53-1 accident conditions by providing post-accident grab sample of the reactor coolant.

4. The PSS supports the CNTS during nonsafety-related Test #53-2 normal operations by providing sampling of containment gas and analysis of hydrogen and oxygen concentration in containment.

5. The PSS supports the condensate nonsafety-related Test #53-3 and FWS by providing sampling and analysis of condensate and feedwater.

6. The PSS supports the MSS by nonsafety-related Test #53-3 providing sampling and analysis of main steam.

7. The PSS system supports the ABS by nonsafety-related Test #53-3 providing sampling and analysis of the auxiliary boiler steam and feedwater.

8. PSS supports the CNTS during nonsafety-related Test #53-2 accident condition by providing sampling of containment gas and analysis of hydrogen and oxygen concentration in containment to respond to emergencies.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each PSS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each PSS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each PSS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

Tier 2 14.2-124 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-53: Process Sampling System Test # 53 (Continued) iv. Verify each PSS system instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each PSS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Test #53-1 Test Objective Test Method Acceptance Criteria Verify sampling capability of the primary i. The NPM is in hot functional testing i. The PSS analysis panel instruments sampling points. with the reactor coolant system at provide indication of the water normal operating pressure and the analysis.

maximum operating temperature ii. The primary sampling ion achievable by heating the RCS with chromatography unit monitors for the MHS. the programmed ion.

The RCS supply and discharge flow is iii. An RCS injection flow grab sample is in service. successfully obtained.

Align the CVCS and PSS to provide iv. An RCS discharge flow grab sample continuous sampling flow to the PSS is successfully obtained.

analysis panel. v. A CVCS demineralizer discharge flow ii. The RCS discharge line is in service. grab sample is successfully Align the RCS and PSS to provide obtained.

sampling flow to the primary sampling ion chromatography units.

iii. Open the PSS grab sample panel manual valve to obtain an RCS injection flow pressurized grab sample.

iv. Open the PSS grab sample panel manual valve to obtain an RCS discharge flow pressurized grab sample.

v. Open the PSS grab sample panel manual valve to obtain a CVC demineralizer discharge flow pressurized grab sample.

Tier 2 14.2-125 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-53: Process Sampling System Test # 53 (Continued)

System Level Test #53-2 Test Objective Test Method Acceptance Criteria Verify sampling capability of the The NPM is in hot functional testing with The PSS containment gas sample panel containment sampling points. the RCS at normal operating pressure instruments provide indication of the and the maximum operating gas analysis.

temperature achievable by heating the RCS with the MHS.

The CES is in service.

Align the CES and PSS to provide continuous sampling flow to the PSS containment gas sample panel.

System Level Test #53-3 Test Objective Test Method Acceptance Criteria Verify sampling capability of the i. The NPM is in hot functional testing i. The PSS secondary sampling system secondary sampling points. with the RCS at normal operating feedwater/main steam sample panel pressure and the maximum instruments provide indication of operating temperature achievable by the water and steam analysis.

heating the RCS with the MHS. ii. The feedwater/main steam ion The FWS and MSS are in service. chromatography analysis panel Align the FWS, MSS, and PSS to monitors the programmed ion.

provide continuous sampling flow to iii. The feedwater/main steam ion the PSS secondary sampling system chromatography analysis panel feedwater/main steam sample panel. monitors the programmed ion.

ii. Open the manual feedwater/main iv. The feedwater/main steam ion steam ion chromatography analysis chromatography analysis panel panel valve to obtain a feedwater to monitors the programmed ion.

SG sample. v. The feedwater/main steam ion iii. Open the manual feedwater/main chromatography analysis panel steam ion chromatography analysis monitors the programmed ion.

panel valve to obtain an SG-1 steam sample.

iv. Open the manual feedwater/main steam ion chromatography analysis panel valve to obtain a SG-2 steam sample.

v. Open the manual feedwater/main steam ion chromatography analysis panel valve to obtain a condensate pump discharge sample.

Tier 2 14.2-126 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-54: 13.8kV and Switchyard System Test # 54 Preoperational test is required to be performed once for the 6A 13.8 kV and switchyard system (EHVS) and once for the 6B EHVS.

The EHVS is described in Sections 8.1.2.1 and 8.3.1.1, and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The EHVS supports the EMVS by nonsafety-related Component level tests providing electrical power.

2. The EHVS supports the TGS by nonsafety-related Component level tests providing electrical protection and control.

3. The EHVS supports the BPSS by nonsafety-related Component level tests providing electrical protection and control to the auxiliary AC power source.

Prerequisites

i. Verify an instrument calibration has been performed on all EHVS instruments that provide information signals to the plant control system (PCS) for the bus and main power transformer under test.

ii. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

iii. Verify all protective devices associated with the EHVS bus and main power transformer under test is tested before that bus is energized Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each EHVS breaker can be Operate each breaker from the local MCR display and local, visual operated locally. control panel while the breaker is in the observation indicate each breaker test position. opens and closes.

ii. Verify each EHVS breaker can be Operate each breaker from the MCR MCR display and local, visual operated remotely. while the breaker is in the test position. observation indicate each breaker opens and closes.

iii. Verify each EHVS breaker trips on its Simulate each fault condition for a MCR display and local, visual fault conditions. breaker when the breaker is in the test observation indicate each breaker position. opens on each fault condition.

iv. Verify each EHVS bus can be powered Energize each EHVS bus from its main Bus voltage is within design limits.

by offsite power via its main power power transformer.

transformer.

(Test not required if an offsite power system is not provided.)

v. Verify each EHVS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each EHVS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-127 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-54: 13.8kV and Switchyard System Test # 54 (Continued)

System Level Tests None Tier 2 14.2-128 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-55: Medium Voltage AC Electrical Distribution System Test # 55 Preoperational test is required to be performed once for the 6A medium voltage AC electrical distribution system (EMVS) and once for the 6B EMVS. The testing of each EMVS bus which provides power to 00 loads (common system loads) is performed with NPM number 1 EMVS loads.

The EMVS is described in Sections 8.1.2.1 and 8.3.1.1 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The EMVS supports the ELVS by nonsafety-related component-level tests providing electrical power.

2. The EMVS supports the circulating nonsafety-related component-level tests water system (CWS) by providing electrical power to loads.

3. The EMVS supports the CHWS by nonsafety-related component-level tests providing electrical power to loads.

4. The EMVS supports the SCWS by nonsafety-related component-level tests providing electrical power to loads.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify all protective devices associated with the EMVS bus and unit auxiliary transformer under test are tested before that bus is energized. Approved test records indicate each protective device has been calibrated within its required test interval.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each EMVS breaker can be Operate each breaker from the local MCR display and local, visual operated locally. control panel while the breaker is in the observation indicate each breaker test position. opens and closes.

ii. Verify each EMVS breaker can be Operate each breaker from the MCR MCR display and local, visual operated remotely. while the breaker is in the test position. observation indicate each breaker opens and closes.

iii. Verify each EMVS breaker trips on its Simulate each fault condition for a MCR display and local, visual fault conditions. breaker when the breaker is in the test observation indicate each breaker position. opens on each fault condition.

iv. Verify each EMVS bus can be Energize each EMVS bus from its unit Bus voltage is within design limits.

powered via its unit auxiliary auxiliary transformer and its adjacent transformer and adjacent EMVS bus. EMV bus.

(Test not required if an offsite power system is not provided.)

v. Verify each EMVS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each EMVS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-129 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-55: Medium Voltage AC Electrical Distribution System Test # 55 (Continued) vi. Verify the automatic transfer of each Simulate all conditions that require an MCR display and local, visual EMVS bus to its adjacent EMVS bus. automatic bus transfer to an adjacent observation indicate the required tie bus. breaker from the adjacent bus closes.

This test may be performed with the EMVS bus energized or deenergized.

System Level Tests None Tier 2 14.2-130 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-56: Low Voltage AC Electrical Distribution System Test # 56 Preoperational test is required to be performed in support of the testing of each NPM. The testing of each ELVS bus, which provides power to common system loads and 6A loads, is performed with NPM number 1 ELVS loads.

The ELVS is described in Section 8.1.2.1 and 8.3.1.1, and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The ELVS provides AC power to nonsafety-related component-level tests system loads via ELVS buses.

2. The ELVS supports the EMVS by nonsafety-related component-level tests providing AC power to the systems auxiliary equipment.

3. The ELVS supports the high voltage nonsafety-related component-level tests AC electrical distribution system (EHVS) by providing AC power to the systems auxiliary equipment.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify all protective devices associated with the ELVS bus and unit auxiliary transformer under test are tested before that bus is energized.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each ELVS breaker can be Operate each breaker from the local MCR display and local, visual operated locally. control panel while the breaker is in the observation indicate each breaker test position. opens and closes.

ii. Verify each ELVS breaker can be Operate each breaker from the MCR MCR display and local, visual operated remotely. while the breaker is in the test position. observation indicate each breaker opens and closes.

iii. Verify each ELVS breaker trips on its Simulate each fault condition for a MCR display and local, visual fault conditions. breaker when the breaker is in the test observation indicate each breaker position. opens on each fault condition.

iv. Verify each ELVS bus can be powered Energize each ELVS bus from its unit Bus voltage is within design limits.

by offsite power via its unit auxiliary auxiliary transformer.

transformer. (Test not required if an offsite power system is not provided.)

v. Verify automatic bus transfer of each Perform the following test for each of the The ELVS bus tie breaker is closed and ELVS bus. ELVS buses. the bus voltage is within design limits.

Open the ELVS supply breaker to a given ELVS bus.

vi. Verify each ELVS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each ELVS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-131 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-56: Low Voltage AC Electrical Distribution System Test # 56 (Continued)

System Level Tests None Tier 2 14.2-132 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-57: Highly Reliable DC Power System Test # 57 Preoperational test is required to be performed for each NPM.

The EDSS is described in Sections 8.1.2.2, 8.1.4.2 and 8.3.2.1.1, and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

The highly reliable DC power system nonsafety-related The Site Acceptance Test criteria in the (EDSS) supports the following systems by prerequisites satisfies the functional providing DC electrical power. verification.

MPS

neutron monitoring system (NMS)

fixed area radiation monitoring system (RMS)

plant lighting system (PLS)

PPS

safety display information system

CRVS EDS system functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. EDSS supports the MPS by providing nonsafety-related Reference 14.2-66 EDSS module-specific operating Component level test parameter information signals.

2. EDSS supports the PPS by providing nonsafety-related Reference 14.2-66 EDSS common operating parameter Component level test information signals.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify a valve-regulated lead-acid battery acceptance tests has been performed on all EDSS batteries to confirm battery capacity in accordance with IEEE Standard 1188 Sections 6 and 7.

iii. Verify battery charger performance testing has been completed by the manufacturer or a site acceptance test has been completed in accordance with manufacturer instructions.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify on loss of power each EDSS De-energize the ELVS motor control The battery charger ELVS input breaker battery charger ELVS input breaker center feed to a EDSS battery charger. is open.

automatically opens. Repeat test for remaining EDSS battery chargers.

ii. Verify each EDSS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each EDSS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Tier 2 14.2-133 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-57: Highly Reliable DC Power System Test # 57 (Continued)

System Level Tests None Tier 2 14.2-134 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-58: Normal DC Power System Test # 58 Preoperational test is required to be performed for each NPM.

EDNS is described in Sections 8.1.2.2, 8.1.4.2 and 8.3.2.1.2 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

The normal DC power system supports nonsafety-related Functions verified by prerequisite and the following systems by providing DC component test.

electrical power for instrumentation and control power.

13.8KV and SWYD

medium voltage AC electrical distribution system

low voltage AC electrical distribution system

PCS

MCS

communication systems

RMS

meteorological and environmental monitoring system

FDS

FPS

Seismic monitoring system (SMS)

TGS

Turbine Building HVAC system (TBVS)

RBVS

plant-wide video monitoring system

CRDS Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii Verify a valve-regulated lead-acid battery acceptance tests has been performed on all EDN batteries to confirm battery capacity in accordance with IEEE Standard 1188 Sections 6 and 7.

iii. Verify battery charger performance testing has been completed by the manufacturer or a site acceptance test has been completed in accordance with manufacturer instructions.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify on loss of power each EDNS De-energize the ELVS motor control The battery charger ELVS input breaker battery charger ELVS input breaker center feed to a EDNS battery charger. is open.

automatically opens. Repeat test for remaining EDNS battery chargers.

Tier 2 14.2-135 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-58: Normal DC Power System Test # 58 (Continued) ii. Verify each EDNS instrument is Initiate a single real or simulated i. The instrument signal is displayed monitored in the MCR and the RSS, if instrument signal from each EDNS on an MCR workstation or recorded the signal is designed to be displayed transmitter. by the applicable control system in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

System Level Tests None Tier 2 14.2-136 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-59: Backup Power Supply Test # 59 Preoperational test is required to be performed for 6A NPMs and 6B NPMs.

The BPSS is described in Sections 8.1.2.1 and 8.1.2.2 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The BPSS supports ELVS by providing nonsafety-related Test #59-1 diesel generator backup electrical power to 480V B-6000 motor control centers.

2. The BPSS supports ELVS by providing nonsafety-related Test #59-1 diesel generator backup electrical power to the operation selected RXB exhaust fan A and B.

3. The BPSS supports ELVS by providing nonsafety-related Test #59-1 diesel generator backup electrical power to the operation selected normal DC power system.

Prerequisites

i. Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

ii. Verify all protective devices associated with the BPSS diesel generators have been tested before performing this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each BPSS remotely-operated Operate each valve from the MCR and MCR display and local, visual valve can be operated remotely. local control panel (if design has local observation indicate each valve fully valve control). opens and fully closes.

ii. Verify each BPSS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of air. Isolate and vent air to the valve. observation indicate each valve fails to its safe position.

iii. Verify each BPSS air-operated valve Place each valve in its non-safe position. MCR display and local, visual fails to its safe position on loss of Isolate electrical power to each air- observation indicate each valve fails to electrical power to its solenoid. operated valve. its safe position.

iii. Verify each BPSS pump can be Align the BPSS to allow for pump MCR display and local, visual started and stopped remotely. operation. observation indicate each pump starts Stop and start each pump from the MCR. and stops.

Audible and visible water hammer are not observed when the pump starts.

iv. Verify the BPSS diesel generator can Align the BPSS to allow for diesel i. and ii.

be started and stopped remotely. generator operation. MCR display and local, visual

i. Start and stop the diesel generator observation indicate the diesel from the MCR. generator started and stopped.

ii. Start and stop the diesel generator locally.

v. Verify the BPSS diesel generator fuel Align the fuel oil transfer pump to MCR display and local, visual oil transfer pump automatically provide oil to the day tank. observation indicate the transfer pump maintains day tank level. Simulate a low level in the day tank. starts.

vi. Verify protective features of the BPSS Align the BPSS to allow for diesel MCR display and local, visual diesel generator. generator operation. Start a diesel observation indicate the diesel generator. generator stops.

Initiate a simulated fault signal for each diesel generator fault condition.

Tier 2 14.2-137 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-59: Backup Power Supply Test # 59 (Continued)

System Level Test #59-1 Test Objective Test Method Acceptance Criteria Verify BPSS diesel generator Align the BPSS to allow for diesel MCR display and local, visual automatically starts and achieves rated generator operation. observation indicate the diesel voltage and frequency. Initiate a real or simulated loss of power generator started and achieved rated signal. voltage and frequency.

Tier 2 14.2-138 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-60: Plant Lighting System Test # 60 Preoperational test is required to be performed once.

The plant lighting system (PLS) is described in Section 9.5.3 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. PLS supports the CRB by providing nonsafety-related component-level test i.

normal lighting.

2. The PLS supports the CRB by nonsafety-related component-level test ii.

providing emergency lighting in the main control room.

3. The PLS supports the RXB by nonsafety-related component-level test i.

providing normal lighting.

4. The PLS supports the RXB by nonsafety-related component-level test ii.

providing emergency lighting for the remote shutdown station.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify the PLS provides normal With normal MCR and RSS lighting in i. a. The PLS provides at least 100 illumination of the MCR and RSS service, measure the light at each MCR foot-candles illumination at the operator workstations, and the MCR and RSS workstation. MCR operator workstations and safety display information panel. at least 50 foot-candles at the MCR auxiliary panels.

[ITAAC 03.08.01]

i. b. The PLS provides at least 100 foot-candles illumination at the RSS operator workstations.

[ITAAC 03.08.01]

ii. The PLS provides emergency With MCR and RSS emergency ii. a. The PLS provides at least 10 illumination of the MCR and RSS illumination in service, measure the light foot-candles of illumination at operator workstations and the MCR at each MCR and RSS workstation and the MCR operator workstations safety display information panel. MCR safety display information panel. and the RSS auxiliary panels.

[ITAAC 03.08.02]

ii. b. The PLS provides at least 10 foot-candles at the RSS operator workstations.

[ITAAC 03.08.02]

iii. Verify the eight-hour battery pack With no AC power available, measure the iii. The required target areas are emergency lighting fixtures provide light at each eight-hour battery pack illuminated to provide at least one illumination for post-fire safe- emergency lighting fixture target area. foot-candle illumination in the areas shutdown activities performed by outside the MCR or RSS where post-operators outside the MCR and RSS. fire safe-shutdown activities are performed.

[ITAAC 03.08.03]

Tier 2 14.2-139 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-60: Plant Lighting System Test # 60 (Continued) iv. Verify each PLS instrument is Initiate a single real or simulated iv. a. The instrument signal is monitored in the MCR and the RSS, if instrument signal from each PLS displayed on an MCR the signal is designed to be displayed transmitter. workstation or recorded by the in the RSS. applicable control system (Test not required if the instrument historian.

calibration verified the MCR and RSS b. The instrument signal is display.) displayed on an RSS workstation or recorded by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

c. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display information monitor if the instrument signal is designed to be displayed on a safety display information monitor.

System Level Tests None Tier 2 14.2-140 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-61: Module Control System Test # 61 Preoperational test is required to be performed as indicated by tests for MCS-controlled systems and systems providing data to the MCS.

The MCS is described in Section 7.0.4.5.

On-site testing of the system is performed by an MCS site acceptance test (SAT).

The MCS is a distributed control system which allows monitoring and control of NPM-specific plant components. The MCS includes all manual controls and visual display units necessary to provide operator interaction with the process control mechanism.

The boundary of the MCS is at the terminations on the MCS hardware. The MCS supplies nonsafety inputs to the human-system interfaces (HSIs) for nonsafety displays in the MCR, the remote shutdown station, and other locations where module control system HSIs are necessary. There are two boundaries between MCS and MPS, the fiber-optic isolated portion and the hard-wired module boundary. The MCS has a direct, bi-directional interface with the PCS.

A complete staging and testing of system hardware and software configurations will be conducted. This factory acceptance testing will be conducted in accordance with a written test procedure for testing the software and hardware of the MCS prior to installation in the plant. Following installation, site acceptance testing shall be completed in accordance with developed procedures to ensure the MCS is installed and fully functional as designed.

To ensure the MCS communicates with module-specific plant components, component-level testing is performed on all systems controlled by MCS to manually operate the associated components from the main control room and remote shutdown station. These component-level tests are described in the test abstracts of the systems that contain the actuated components.

In addition, it is verified that each instrument supplying data to the MCS is component tested in preoperational test abstracts to ensure the signal is displayed in the MCR and RSS if applicable. These component-level tests are described in the test abstracts of the systems that contain the instrument.

System Function System Function Categorization Function Verified by Test #

Verify each MCS-controlled system Initiate a single real or simulated i. The instrument signal is displayed instrument is monitored in the MCR and instrument signal from each MCS- on an MCR workstation or recorded the RSS, if the signal is designed to be controlled system transmitter. by the applicable control system displayed in the RSS. historian.

(Test not required if the instrument ii. The instrument signal is displayed calibration verified the MCR and RSS on an RSS workstation or recorded display.) by the applicable control system historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Prerequisites Prerequisites associated with MCS testing are identified in the test abstracts that contain module-specific components that ensure communication with the MCS.

Component Level Tests None System Level Tests None Tier 2 14.2-141 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-62: Plant Control System Test # 62 Preoperational test is required to be performed as indicated by tests for PCS-controlled systems and systems providing data to the PCS.

The PCS is described in Section 7.0.4.6.

On-site testing of the system is performed by a PCS site acceptance test (SAT).

The PCS is a distributed control system which allows monitoring and control of virtually all module-specific plant components. The PCS includes all manual controls and visual display units necessary to provide operator interaction with the process control mechanism.

The boundary of the PCS is at the terminations on the PCS hardware. The PCS supplies nonsafety inputs to the video display units (VDUs) for nonsafety displays in the MCR, the RSS, and other locations where PCS video display units are necessary. The boundary between the PPS and PCS is at the output connection of the safety-related optical isolators in the PPS, and on the terminals of the equipment interface module for each input from the PCS to the PPS.

The PCS has a direct, bidirectional interface with the MCS. The network interface devices for the PCS domain controller/

historian provide the interface between the human machine interface network layer and the control network layer.

A complete staging and testing of system hardware and software configurations will be conducted. This factory acceptance testing will be conducted in accordance with a written test procedure for testing the software and hardware of the PCS prior to installation in the plant. Following installation, site acceptance testing shall be completed in accordance with developed procedures to ensure the PCS is installed and fully functional as designed.

To ensure the PCS communicates with module-specific plant components, component-level testing is performed on all systems controlled by PCS to manually operate the associated components from the main control room and remote shutdown station. These component-level tests are described in the test abstracts of the systems that contain the actuated components.

In addition, it is verified that each instrument supplying data to the PCS is component tested in preoperational test abstracts to ensure the signal is displayed in the MCR and RSS if applicable. These component-level tests are described in the test abstracts of the systems that contain the instrument.

System Function System Function Categorization Function Verified by Test #

Verify each PCS-controlled system Initiate a single real or simulated i. The instrument signal is displayed instrument is monitored in the main instrument signal from each PCS- on an MCR workstation or recorded control room (MCR) and the remote controlled system transmitter. by the applicable control system shutdown station (RSS), if the signal is historian.

designed to be displayed in the RSS. ii. The instrument signal is displayed (Test not required if the instrument on an RSS workstation or recorded calibration verified the MCR and RSS by the applicable control system display.) historian if the instrument signal is designed to be displayed in the RSS.

iii. The instrument signal is displayed on an MCR module-specific safety display instrument monitor or an MCR common safety display instrument monitor if the instrument signal is designed to be displayed on a safety display instrument monitor.

Prerequisites Prerequisites associated with PCS testing are identified in the test abstracts that contain module-specific components that ensure communication with or are controlled by the PCS.

Component Level Tests None System Level Tests None Tier 2 14.2-142 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Module Protection System Test #63 Preoperational test is required to be performed for each NPM.

The MPS is described in Sections 7.0, 7.1, and 7.2 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

The MPS supports the CNTS by removing safety-related Test #63-4 electrical power to the trip solenoids of Test #63-6 the following CIVs on a containment system isolation actuation signal:

RCS injection containment isolation valves

RCS discharge containment isolation valves

Pressurizer spray containment isolation valves

RPV high point degasification containment isolation valves

Feedwater containment isolation valves

Main steam containment isolation valves

Main steam bypass valves

CES containment isolation valves

RCCWS containment isolation valves

CFDS containment isolation valves The MPS supports the CNTS by removing safety-related Test #63-4 electrical power to the trip solenoids of Test #63-6 the following valves on a DHRS actuation signal.

DHRS actuation valves

Main steam isolation valves

Main steam bypass isolation valves

Feedwater isolation valves The MPS supports theECCS by removing safety-related Test #63-4 electrical power to the trip solenoids of Test #63-6 the following valves on an ECCS actuation signal.

Reactor vent valves

Reactor recirculation valves The MPS supports the Containment safety-related Test #63-4 system by removing electrical power to Test #63-6 the trip solenoids of the following containment isolation valves on a CVCS isolation actuation signal:

RCS injection containment isolation valves

RCS discharge containment isolation valves

PZR pressurizer spray CIVs

RPV high point degasification containment isolation valves Tier 2 14.2-143 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Module Protection System Test #63 (Continued)

The MPS supports the CVCS by removing safety-related Test #63-4 electrical power to the trip solenoids of Test #63-6 the DWS supply isolation valves on a DWS isolation actuation signal The MPS supports the ECCS by removing safety-related Test #63-4 electrical power to the trip solenoids of Test #63-6 the reactor vent valves on an LTOP actuation signal.

The MPS supports the ELVS by removing safety-related Test #63-4 electrical power to the pressurizer Test #63-6 heaters on a pressurizer heater trip actuation signal.

The MPS supports the EDNS by removing safety-related Test #63-4 electrical power to the CRDS for a reactor Test #63-5 trip.

The DHRS supports the RCS by opening safety-related Test #63-6 the DHRS actuation valves on a DHRS actuation signal for DHRS operation.

The CNTS supports the DHRS by closing safety-related Test #63-6 CIVs for the main steam and feedwater systems when actuated by the MPS.

The CNTS supports the RCS by closing safety-related Test #63-6 the CIVs for pressurizer spray, RCS injection, RCS letdown, and RPV high point degasification when actuated by the MPS.

The CNTS supports the RXB by providing safety-related Test #63-6 a barrier to contain mass, energy, and fission product release by closure of the CIVs upon a containment isolation signal.

The ECCS supports the RCS by opening safety-related Test #63-6 the ECCS reactor vent valves and reactor recirculation valves when their respective trip valve is actuated by the MPS.

The ECCS supports the RCS by providing safety-related Test #63-6 recirculated coolant from the containment to the RPV for the removal of core heat.

The ECCS supports the RCS by providing safety-related Test #63-6 LTOP for maintaining the reactor coolant pressure boundary.

The CVCS supports the RCS by isolating safety-related Test #63-6 dilution sources.

The FWS supports the CNTS by providing nonsafety-related Test #63-6 secondary isolation of the feedwater lines.

The MSS supports the CNTS by providing nonsafety-related Test #63-6 secondary isolation of the main steam lines.

Tier 2 14.2-144 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Module Protection System Test #63 (Continued)

The FWS supports the DHRS by nonsafety-related Test #63-6 providing secondary isolation of the feedwater lines, ensuring required boundary conditions for DHRS operation.

The NMS supports the MPS by providing safety-related Test #63-4 neutron flux data for various reactor trips.

ECCS supports MPS by providing nonsafety-related Test #63-1 instrumentation information signals.

The DHRS supports the MPS by safety-related Test #63-1 providing MPS actuation instrument information signals.

The RCS supports the MPS by providing nonsafety-related Test #63-1 instrument information signals.

The RCS supports the MPS by providing safety-related Test #63-1 instrument information signals for LTOP actuation The CVCS supports ECCS valves by nonsafety-related Test #63-6 providing water to reset the ECCS valves.

Prerequisite Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests None System Level Test #63-1 Test Objective Test Method Acceptance Criteria Verify the instrument signals of MPS Table 7.1-2 lists all of sensors which input Each MPS monitored signal is displayed monitored variables are displayed in the to MPS. on an MCR workstation and the module-MCR. This test may be performed concurrently specific safety display instrument panel with safety display and indication system (if designed for safety display instrument (SDIS) test #66 -2 for PAM Type B and display).

Type C testing described in Section 14.2.12.

Inject a single signal as close as practical for each sensor listed in Table 7.1-2 and monitor its response on an MCR workstation and the module-specific safety display instrument panel (if designed for safety display instrument display).

If the sensor signal is designed to be disconnected when the NPM is moved then it will be necessary to test the signal from the sensor to the disconnect and then from the disconnect to the MCR display.

Tier 2 14.2-145 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Module Protection System Test #63 (Continued)

System Level Test #63-2 Test Objective Test Method Acceptance Criteria

i. Verify the reactor trip logic fails to a This test will verify initiation of a trip i. Loss of electrical power in a safe state such that loss of electrical state for MPS separation groups on loss separation group results in a reactor power to an MPS separation group of power to that separation group. trip state for that separation group.

results in a reactor trip state for that Component actuation is not required or [ITAAC 02.05.14]

separation group. verified.

ii. Loss of electrical power in a ii. Verify the ESF logic fails to a safe i. Remove power from one separation separation group results in the state such that loss of electrical group of one reactor trip function predefined state for that separation power to an MPS separation group listed in Table 7.1-3 to provide a trip group.

results in a predefined safe state for state for that separation group. [ITAAC 02.05.15]

that separation group. Repeat tests for all separation groups for all reactor trip functions.

ii. Remove power from one separation group of one ESF actuation function listed in Table 7.1-3 to provide a predefined state for that separation group.

Repeat tests for all separation groups for all ESF actuation functions.

System Level Test #63-3 Test Objective Test Method Acceptance Criteria

i. Verify MPS operating bypass This test will verify operation of each i. a. The operating bypasses are interlocks are automatically operating bypass interlock and automatically established.

established when the associated operating bypass permissive. b. The operating bypasses are interlock condition is satisfied and Component actuation is not required or automatically removed.

automatically removed when the verified.

[ITAAC 02.05.18]

condition is not satisfied. Table 7.1-5 contains the following ii. Verify MPS operating bypasses can ii. a. The operating bypasses can be information:

be manually established when the manually established.

The identification of each operating associated permissive condition is b. The operating bypasses are bypass interlock and operating bypass satisfied and automatically removed automatically removed.

permissive and their logic.

when the condition is not satisfied.

The function of the operating bypass [ITAAC 02.05.19]

iii. Verify MCR alarms when operating i. a. Simulate the logic for an iii. Each established operating bypass is bypasses are established. operating bypass interlock. alarmed in the MCR.

b. Remove the logic. [ITAAC 02.05.22]

Repeat test for all operating bypass interlocks.

ii. a. Simulate the logic for an operating bypass permissive and manually establish the operating bypass.

b. Remove the logic.

Repeat test for all operating bypass permissives.

Tier 2 14.2-146 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Module Protection System Test #63 (Continued)

System Level Test #63-4 Test Objective Test Method Acceptance Criteria

i. Verify the MPS automatically initiates This test verifies initiation of reactor trip i. A reactor trip signal is displayed in a reactor trip signal. signals and ESF actuation signals only. the MCR for all 2 out of 4 logic ii. Verify the MPS automatically initiates Component actuation is not required or combinations of each reactor trip an ESF actuation signal. verified. function.

Test #63-1 is completed in order to use [ITAAC 02.05.08]

the associated test signals. ii. An ESF actuation signal is displayed Real or simulated CNTS level, reactor trip in the MCR for all 2 out of 4 logic breaker position, RCS temperature and combinations each reactor ESF NMS signals may be required to provide actuation function.

the necessary bypass interlock status for [ITAAC 02.05.09]

either the reactor trip or ESF actuation to be available.

i. Initiate an automatic reactor trip signal by simulating a reactor trip function for each function listed in Table 7.1-3.

All combinations of the 2 out of 4 logic are tested for each reactor trip function.

ii. Initiate an automatic ESF actuation signal by simulating an ESF actuation function for each function listed in Table 7.1-4. All combinations of the 2 out of 4 logic must be actuated for each ESF function.

System Level Test #63-5 Test Objective Test Method Acceptance Criteria

i. Verify the MPS manually actuates a This test will verify the automatic and i. The reactor trip breakers open.

reactor trip. manual reactor trips. Reactor trip [ITAAC 02.05.12]

ii. Verify the MPS automatically actuation is verified by reactor trip ii. The reactor trip breakers open.

actuates a reactor trip. breaker actuation. Only one trip function is required to perform the automatic [ITAAC 02.05.10]

reactor trip,

i. Initiate a manual reactor trip from the MCR.

ii. Initiate an automatic reactor trip signal by simulating any single reactor trip function.

Tier 2 14.2-147 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Module Protection System Test #63 (Continued)

System Level Test #63-6 Test 63-6 is performed at hot functional testing concurrently with TGS test #33-1 (reference 14.2.12.33) to allow testing of ESF actuations at normal operating pressure and elevated temperatures. Test #33-1 heats the RCS from ambient conditions to the highest temperature achievable by MHS heating. These hot functional testing conditions provide the highest differential pressure and temperature conditions that can be achieved prior to fuel load.

Test Objective Test Method Acceptance Criteria

i. Verify the MPS can manually actuate Figure 7.1-1 identifies all ESF actuation i. The MPS actuates the ESF equipment ESF equipment from the MCR. signals such as CVCS isolation and CNTS to perform its safety-related function ii. Verify deliberate operator action is isolation. as described in Table 7.

required to return the ESF actuated Table 7.1-4 lists all of the ESF functions. Each ECCS valve opens after receipt equipment to its non-actuated This test will verify the design response of an ESF signal and after RCS position. of ESF actuation signals using both a pressure is decreased to the iii. Verify the MPS can automatically single manual ESF signal and a single ESF threshold pressure for operation of actuate ESF equipment from all ESF function to provide an automatic ESF the inadvertent actuation block actuation signals. actuation signal. All manual and described in described in Section automatic ESF actuation signals are 6.3.2.2.

tested. [ITAAC 02.01.13]

The RCS is at normal operating pressure [ITAAC 02.01.14]

supplying bypass steam to the [ITAAC 02.01.15]

condenser.

[ITAAC 02.01.18]

i. Initiate a manual ESF actuation signal

[ITAAC 02.01.19]

from the MCR.

[ITAAC 02.01.20]

ii. a. Attempt to operate the actuated ESF equipment from the MCR. [ITAAC 02.05.13]

b. Remove the manual ESF [ITAAC 02.05.16]

actuation signal and attempt to ii. a. The actuated equipment cannot operate the actuated ESF be operated from the MCR.

equipment from the MCR. b. The actuated equipment cannot

c. Use the MCR enable nonsafety be operated from the MCR.

control switch to allow operation c. The ESF equipment can be of the ESF actuated equipment operated from the MCR.

from the MCR.

[ITAAC 02.01.13]

Repeat for all MCR manual ESF actuations. [ITAAC 02.01.14]

iii. Initiate an automatic ESF actuation [ITAAC 02.01.15]

signal. The test may be performed [ITAAC 02.05.16]

with the RCS at ambient conditions. iii. The MPS automatically actuates the Repeat for all ESF actuation signals. ESF equipment to perform its safety-related function as described in Table 7.1-4.

[ITAAC 02.01.13]

[ITAAC 02.01.14]

[ITAAC 02.01.15]

[ITAAC 02.01.18]

[ITAAC 02.01.20]

[ITAAC 02.05.11]

[ITAAC 02.05.16]

Tier 2 14.2-148 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Module Protection System Test #63 (Continued)

System Level Test #63-7 Test #63-7 is performed concurrently with Test #63-6 which operates all of the ESF actuation valves during hot functional testing.

Test #63-7 records the stroke times of containment isolation valves (CIVs) as they travel to their ESF-actuated position with the RCS pressure at normal operating pressure.

Test Objective Test Method Acceptance Criteria Verify the CIVs operate to satisfy their Table 6.2-10 contains the design closure i. Each containment isolation valve ESF-actuated design stroke time. time for containment isolation valves. travels from fully open to fully closed in less than or equal to the time listed in Section 6.2.4.3 after receipt of a Time the operation of all CIVs as they containment isolation signal.

actuate to their ESF position during the manual ESF actuation testing in Test #63- [ITAAC 02.01.08]

6. [ITAAC 02.05.17]

System Level Test #63-8 This test will verify the time response of MPS reactor trip and ESF actuation signals. The reactor trip test verifies response time through reactor trip breaker actuation. The ESF response time is tested through the de-energization of the associated solenoid valve or the opening of the pressurizer heater supply breaker. ESF valve response times are tested in Test #63-7.

Test Objective Test Method Acceptance Criteria Verify the MPS response times from Section 7.1.4 contains a description of The MPS reactor trip functions listed in sensor output through: design basis event actuation delays Table 7.1-3 and ESF functions listed in

i. reactor trip breaker actuation for the assumed in the plant safety analysis and Table 7.1-4 have response times that are reactor trip function. listed in Table 7.1-6. The actuation delays less than or equal to the design basis do not include ESF actuated component safety analysis response time ii. de-energization of the associated delays for actuated valves. assumptions in Table 7.1-6.

solenoid valve for ESF-actuated valves.

iii. opening of the pressurizer heater Perform a time response test for the [ITAAC 02.05.17]

supply breaker for the pressurizer actuation signals listed in Table 7.1-6.

heater trip.

Response time testing for ESF actuated CNTS, DHRS, ECCS and DWS valves are found in Test #63-7.

System Level Test #63-9 Test Objective Test Method Acceptance Criteria Verify protective measures are provided Section 7.2.9.1 provides the manual All actions described in Section 7.2.9.1 to restrict modifications to the MPS actions required to modify tunable are required to modify tunable tunable parameters. parameters. parameters.

A test will be performed to verify that all [ITAAC 02.05.02]

manual actions described in Section 7.2.9.1 are required to modify tunable parameters.

Tier 2 14.2-149 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Module Protection System Test #63 (Continued)

System Level Test #63-10 Test Objective Test Method Acceptance Criteria

i. Verify the MPS is capable of Section 7.2.4.2 discusses the operation of i. a. The SFM out of service provides a performing its safety-related the MPS maintenance bypass operation no trip to the respective functions when one of its separation for the MPS safety function module scheduling and voting module.

groups is placed in maintenance (SFM). b. There is no change to the 2 out of bypass. 4 voting logic for the separation ii. Verify MPS maintenance bypasses Place an SFM in maintenance bypass by group.

are indicated in the main control using the out of service and trip/bypass [ITAAC 02.05.21]

room. switches associated with the SFM. ii. The inoperable status of the SFM is provided in the MCR.

Repeat tests for all SFMs. [ITAAC 02.05.23]

System Level Test #63-11 Test Objective Test Method Acceptance Criteria Verify the controls located on the i. A test will be performed to verify the i. Water is added to the RCS.

operator workstations in the MCR CVCS can add water to the RCS after ii. Water is added to containment.

operate to perform important human a containment isolation signal using

[ITAAC 02.05.20]

actions. the O-1 override switch and MCR controls. [ITAAC 02.05.26]

ii. A test will be performed to verify the (i. and ii.)

CFDS can add water to containment after a containment isolation signal using the O-1 override switch and MCR controls.

Tier 2 14.2-150 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-64: Plant Protection System Test # 64 Preoperational test is required to be performed once.

The PPS is described in Section 7.0.4.3. PPS functions are not verified by PPS tests. PPS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. The PPS supports the CRVS by nonsafety-related CRHS Test #18-1 providing actuation and control signals to the CRE isolation dampers.

2. The PPS supports the control room nonsafety-related CRHS Test #18-1 habitability system (CRHS) by providing actuation and control signals.

3. The PPS supports the CRVS by nonsafety-related CRVS Test #19-3 providing actuation and control signals to the outside air isolation dampers.

Prerequisites Prerequisites associated with PPS testing are identified in the referenced test abstract cited under the Function Verified by Test # heading.

Component Level Tests None System Level Tests None Tier 2 14.2-151 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-65: Neutron Monitoring System Test # 65 Preoperational test is required to be performed for each NPM.

The NMS is described in Section 7.0.4.2 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The NMS supports the MPS by safety-related Test #63-4 providing neutron flux data for various reactor trips.

2. The NMS supports the MPS by nonsafety-related Test #66-2 providing information signals for PAM.

3. The NMS supports the MPS by nonsafety-related Test #66-2 providing information signals for PAM during CNV flooded conditions.

Prerequisites Prerequisites associated with NMS testing are identified in the referenced test abstract cited under the Function Verified by Test # heading.

Component Level Tests None System Level Tests None Tier 2 14.2-152 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-66: Safety Display and Indication Test # 66 Test #66 Component-level testing for the module-specific SDIS is required to be performed for each NPM.

Test # 66 Component-level testing for the common SDIS is required to be performed once.

Test #66-1 System-level testing for the module-specific SDIS is required to be performed for each NPM to verify proper trending of RCS pressure and temperature.

Test #66-2 System-level testing for the module-specific SDIS is required to be performed for each NPM to verify PAM variables are displayed and alarms retrieved.

SDIS is described in Section 7.0.4.4 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The SDIS actively supports the CRB nonsafety-related i. Module-specific SDIS component-by providing the main control room level tests accident monitoring plant ii. Common SDIS component-level conditions. tests iii. Test #66-1 iv. Test #66-2

2. The SDIS actively supports the PCS by nonsafety-related i. Module-specific SDIS component-providing plant status and indication level tests data to the plant data historian. ii. Common SDI component-level tests iii. Test #66-1 iv. Test #66-2

3. The ICIS supports the MPS by nonsafety-related Test #66-2 providing RXC temperature information.

4. The ECCS supports MPS by providing nonsafety-related Test #66-2 PAM instrument information signals.

5. The RCS supports the MPS by nonsafety-related Test #66-2 providing PAM instrument information signals.

6. The CNTS supports the MPS by nonsafety-related Test #66-2 providing PAM information signals.

7. The RMS supports the RXB by nonsafety-related Test #66-2 monitoring radiation levels in the building in proximity of the bioshield.

8. The NMS supports the MPS by nonsafety-related Test #66-2 providing information signals for PAM.

9. The NMS supports the MPS by nonsafety-related Test #66-2 providing information signals for PAM during CNV flooded conditions.

10. The DHRS supports the MPS by nonsafety-related Test #66-2 providing PAM instrument information signals.

11. The EDSS supports the PPS by nonsafety-related Component-level test: Common SDIS providing common EDSS operating test iii.

parameter information signals.

12. The EDSS supports the MPS by nonsafety-related Component-level test: Module-Specific providing module-specific EDSS SDIS test iii.

operating parameter information signals.

Tier 2 14.2-153 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-66: Safety Display and Indication Test # 66 (Continued)

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Note:

Testing of PAM Type B and Type C displays and alarms is performed in Test #66-1.

Note:

Testing of NPM level, pressure, and temperature and flow instruments is performed in Test #66-2.

Component Level Tests: Common SDIS Test Test Objective Test Method Acceptance Criteria

i. Verify the proper valve position Open and close each valve listed in Table The valve opens and closes as indicated indication for each valve that 7.1-8. by a common SDIS display and an MCR provides input to the PPS. workstation display.

ii. Verify radiation monitor indication is Provide a simulated signal for each EDSS The radiation signal is displayed by a obtained in the MCR for each and ELVS voltmeter monitored by PPS common SDIS display and an MCR radiation monitor that provides input listed in Table 7.1-8. workstation.

to the PPS.

iii. Verify EDSS and ELVS voltage Provide a simulated signal for each EDSS The voltage signal is displayed by a indication is obtained in the MCR for and ELVS voltmeter monitored by PPS common SDIS display and an MCR voltmeters that provide input to the listed in Table 7.1-8. workstation.

PPS.

iv. Verify instrument indication is Provide a simulated signal for each The instrument signal is displayed by a obtained in the MCR for instruments instrument monitored by PPS listed in common SDIS display and an MCR that provide input to the PPS. Table 7.1-8. workstation.

Component Level Tests: Module Specific SDI Test Test Objective Test Method Acceptance Criteria

i. Verify the proper valve position i. With the NPM assembled, open and i. The valves open and close as indication for each ESF valves that close the valves listed in Table 7.1-2. indicated by a module-specific SDIS provide input to MPS. ii. Provide a real or simulated signal for display and an MCR workstation each reactor safety valve position display.

(Table 7.1-2). ii. The valve opens and closes as indicated by a module-specific SDIS display and an MCR workstation display.

ii. Verify radiation monitor indication is Provide a simulated signal for each The radiation monitor signal is displayed obtained in the MCR for each radiation monitor monitored by MPS by a module-specific SDIS display and radiation monitor that provides input listed in Table 7.1-2. an MCR workstation.

to the MPS.

iii. Verify EDSS and ELVS voltage Provide a simulated signal for each EDSS The voltage signal is displayed by a indication is obtained in the MCR for and ELVS voltmeter monitored by MPS module-specific SDIS display and an each voltmeter that provide input to (Table 7.1-2). MCR workstation.

the MPS.

iv. Verify neutron flux indication is Provide a simulated signal for each The neutron flux signal is displayed by a obtained in the MCR for each neutron flux instrument monitored by module-specific SDIS display and an radiation monitor that provides input MPS (Table 7.1-2). MCR workstation display.

to the MPS.

v. Verify a neutron flux instrument fault Provide a simulated signal for each The neutron flux instrument fault is indication is obtained in the MCR for neutron flux instrument fault monitored displayed by a module-specific SDIS each signal that provides input to the by MPS (Table 7.1-2). display and an MCR workstation display.

MPS.

Tier 2 14.2-154 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-66: Safety Display and Indication Test # 66 (Continued)

System Level Test #66-1 Test 66-1 is conducted concurrently with TGS test# 33-1 which warms the RCS from ambient conditions to the highest temperature achievable by MHS heating.

Test Objective Test Method Acceptance Criteria Verify that the output signals from the Increase RCS temperature from ambient Trended data shows agreement NPM level, pressure, temperature and to the highest temperature achievable between the two divisional instruments flow instruments listed in Table 7.1-2 by MHS heating. or the four safety group instruments properly trend while increasing RCS Using the MCS historian record the monitoring the same variable.

temperature and pressure. engineering values for the output of the Note: This is not a verification of instruments described in the test instrument calibrations. objective. Record data at approximately 50 °F intervals from ambient temperature to the maximum RCS temperature.

Note: Instrument signals are provided to the module-specific SDIS display and the main control room workstations.

System Level Test #66-2 Test Objective Test Method Acceptance Criteria

i. Verify PAM Type B and C variables are i. Simulate an injection signal for the i. The PAM Type B and C variables displayed on the module-specific PAM Type B and C variables listed in listed in Table 7.1-7 are retrieved SDIS displays in the MCR. Table 7.1-7. and displayed on the SDI displays in ii. Verify alarms associated with PAM ii. Increase or decrease a simulated the MCR.

Type B and C variables are retrieved injection signal for the PAM Type B [ITAAC 02.05.25]

in the MCR. and C variables listed in Table 7.1-7 to iii. Verify module-specific PAM Type D obtain its associated alarm. ii. The alarms associated with the PAM variables are displayed on the iii. Simulate an injection signal for the Type B and C variables listed in Table module-specific SDIS displays in the PAM Type D variables listed in Table 7.1-7.are retrieved and displayed on MCR. 7.1-7 the SDI displays in the MCR.

iii. The PAM Type D variables listed in Table 7.1-7 are retrieved and displayed on the SDIS displays in the MCR.

Tier 2 14.2-155 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-67: Fixed Area Radiation Monitoring System Test # 67 Preoperational test is required to be performed once.

The fixed-area radiation monitoring system is described in Section 12.3.4 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The fixed-area radiation monitoring nonsafety-related Component-level test system supports the following buildings by monitoring radiation levels:

ANB

CRB

RWB

TGB

RXB RMS function verified by another test is:

System Function System Function Categorization Function Verified by Test #

The RMS supports the RXB by monitoring nonsafety-related Test #66-2 radiation levels in the building in proximity of the bioshield.

Prerequisites Verify an instrument calibration has been completed, with approved records and within all calibration due dates, for all instruments required to perform this test.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify each fixed airborne radiation Actuate the check source on a fixed MCR display and local, visual monitors response to a known airborne radiation monitor listed in Table observation indicate the following:

source. 12.3-10. i. The main control room audible and Repeat test for the remainder of fixed visual alarms are received.

airborne radiation monitors. ii. The local readout, audible alarm and visual alarm are received.

ii. Verify each fixed area radiation Actuate the check source on a fixed area MCR display and local, visual monitors response to a known radiation monitor listed in Table 12.3-11. observation indicate the following:

source. Repeat test for the remainder of fixed i. The main control room audible and area radiation monitors. visual alarms are received.

ii. The local readout, audible alarm and visual alarm are received.

System Level Tests None Tier 2 14.2-156 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-68: Communication System Test # 68 Preoperational test is required to be performed after construction turnover of the communication system (COMS).

The COMS is described in Section 9.5.2 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

The COMS supports the following nonsafety-related Component-level tests i. through iv.

locations by providing voice and data communications within the building and surrounding areas.

RXB

TGB

RWB

CRB

Security Buildings

ANB

Diesel Generator Building

Administrative and Training Building

Central Utility Building

Warehouse Building

Fire Water Building Prerequisites

i. Required communications system site acceptance tests have been completed and approved.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify plant public address/general Station test personnel in each required i. The test announcement is heard at alarm (PA/GA) system can be heard test area of the plant to monitor the PA/ each test site.

throughout the plant site. GA system. ii. The test emergency alarm is heard at

i. Use the public address to provide a each test site.

test announcement.

ii. Use the general alarm system to provide a test alarm.

ii. Verify plant radio communications Station test personnel in each required The plant radio communication is can be heard throughout the plant test area of the plant to communicate obtained at each test site.

site. using plant radios.

iii. Verify the sound powered telephone Test each sound powered telephone. All channels of each sound powered system can be used for voice telephone can be used to communicate communication. with another sound powered telephone.

iv. Verify wireless communication Station test personnel in each required The voice and data communication is throughout the plant site. test area of the plant to communicate obtained at each test site.

using voice and data communication.

v. Verify the central alarm station is Test the conventional (landline) service The conventional service connects with equipped with a conventional from the central alarm station to the the control room and the local law (landline) telephone service which control room and local law enforcement enforcement authorities.

can be used to communicate with authorities.

the control room and local law [ITAAC 03.16.11]

enforcement authorities.

vi. Verify that plant radio Test communications with the plant The radios will provide continuous communications maintains radio system in areas described in the communications in all test areas.

continuous communications physical protection program boundaries between the central alarm station and areas described in the contingency [ITAAC 03.16.12]

and on-duty watchmen, armed response event areas.

security officers, armed responders, or other security personnel who have responsibilities within the physical protection program and during contingency response events.

Tier 2 14.2-157 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-68: Communication System Test # 68 (Continued) vii. Verify all nonportable Remove normal power from the central The nonportable communication communication devices (including alarm station nonportable devices establish connections with the conventional telephone systems) in communication devices. normal power removed.

the central alarm station remain operable during the loss of normal [ITAAC 03.16.13]

power.

System Level Tests None Tier 2 14.2-158 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-69: Seismic Monitoring System Test # 69 The SMS is described in Section 3.7.4. The SMS is a site-specific system, and the SMS design is the responsibility of the COL applicant as indicated by COL item 3.7-1.

COL Item 14.2-6: A COL applicant that references the NuScale Power Plant design certification will provide a test abstract for the seismic monitoring system pre-operational testing.

System Function System Function Categorization Function Verified by Test #

As described in Section 3.7.4 nonsafety-related Provided by COL applicant Prerequisites Provided by COL applicant Component Level Tests Provided by COL applicant System Level Tests Provided by COL applicant Tier 2 14.2-159 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-70: Hot Functional Testing Test # 70 Preoperational testing is required to be performed once for each NPM.

The following identifies the tests employed in support of the performance of hot functional testing.

Hot Functional Testing Tests Test Objective Verified by Test # Tested Function Categorization CES i. Verifies the automatic i. CE Test #41-1 nonsafety-related operation of the CES to ii. CE Test #41-2 establish and maintain iii. CE Test #41-3 design vacuum for the CNV.

ii. Verify radiation isolation on high radiation level in the ABS.

iii. Verifies the CES supports RCS leakage detection.

CFD i. Verifies the CFDS can i. CFDS Test #42-1 nonsafety-related automatically drain the ii. CFDS Test #42-2 CNTS. This test may be iii. CFDS Test #42-3 completed as a iv. CFDS Test #42-4 prerequisite to hot functional testing because iv. CFDS Test #42-5 the completion of the test does not require elevated water temperatures.

ii. Verifies the CFDS can automatically flood the CNTS. This test may be completed as a prerequisite to hot functional testing because the completion of the test does not require elevated water temperatures.

iii. Verifies the CFDS can provide borated water to the RCS during a beyond design basis accident.

(Important human action).

iv. Verifies the CFDS responds to high radiation conditions.

CVC i. Verifies CVCS automatic i. CVCS Test #38-1 nonsafety-related makeup to maintain ii. CVCS Test #38-2 pressurizer level. iii. CVCS Test #38-3 ii. Verifies automatic pressurizer pressure control.

iii. Verifies CVCS automatic boration and dilution of the RCS.

Tier 2 14.2-160 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-70: Hot Functional Testing Test # 70 (Continued)

ECCS Each ECCS valve opens after i. MPS Test #63-6 nonsafety-related receipt of an ESF signal and after RCS pressure is decreased to the threshold pressure for operation of the inadvertent actuation block.

FWS i. Verifies the FWS i. Test TG #33-1 nonsafety-related automatically controls flow ii. Test TG #33-1 to the SGs to maintain SG inventory.

ii. Verifies the FWS automatically cools the turbine generator bypass steam flow in the main steam desuperheater.

ICIS Verifies proper temperature Test ICI #49-1 nonsafety-related indication is obtained from the ICIS thermocouples.

MHS i. Verifies the MHS is capable i. TG Test #33-1 nonsafety-related of heating the RCS to a ii. TG Test #33-1 temperature sufficient to iii. TG Test #33-1 obtain criticality.

ii. Verifies the MHS is capable of heating the RCS to establish natural circulation flow sufficient to obtain criticality.

iii. Verifies a local grab sample can be obtained from an MHS grab sample device indicated on the MHS piping and instrumentation diagram.

MPS i. Verifies design responses i. Test #63-6 safety-related to manual ESF signals. ii. Test #63-7 ii. Verifies containment iii. Test #63-6 isolation valves closure iv. Test #63-4 times.

v. Test #63-5 iii. Verifies design responses vi. Test #63-6 to automatic ESF signals.

iv. Verifies design responses to automatic reactor trip signals.

v. Verifies automatic enabling and reset of operational bypasses.

vi. Verifies the ECCS valves closes when the CVCS provides water to reset the ECCS valves.

Tier 2 14.2-161 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-70: Hot Functional Testing Test # 70 (Continued)

PSS i. Verifies sampling capability Test #53-1 nonsafety-related of the primary sampling Test #53-2 points. Test #53-3 ii. Verifies sampling capability of the containment sampling points.

ii. Verifies sampling capability of the secondary sampling points.

SDIS Verify that the output signals Test #66-1 nonsafety-related from the NPM level, pressure, temperature, and flow instruments listed in Table 7.1-2 properly trend while increasing RCS temperature and pressure.

TG i. Verifies the TGS i. Test #33-1 nonsafety-related automatically controls ii. Test #33-2 turbine bypass flow to the main condenser.

ii. Verifies the turbine generator can obtain synchronous speed.

Prerequisites Prerequisites associated with performing hot functional testing are identified in the referenced test abstract cited under the Verified by Test # heading.

Tier 2 14.2-162 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-71: Module Assembly Equipment Bolting Test # 71 Preoperational test is required to be performed once.

The MAE bolting is described in Section 9.1.5. MAE bolting functions are not verified by MAE bolting tests. MAE bolting functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. MAE bolting supports the CNTS nonsafety-related Test #52-2 actively by providing material handling to allow for disassembly and reassembly of the CNV lower flange.

2. MAE bolting supports the RPV nonsafety-related Test #52-2 actively by providing material handling to allow for disassembly and reassembly of the RPV lower flange.

Prerequisites Prerequisites associated with MAE bolting testing are identified in the referenced test abstract cited under the Function Verified by Test # heading.

Component Level Tests None System-Level Tests None Tier 2 14.2-163 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-72: Steam Generator Flow-Induced Vibration Test # 72 Validation testing is performed at test facilities as separate effects tests on prototypic steam generator tube columns and steam generator inlet flow restrictors per Table 4-1 of TR-0716-50439. There are no preoperational tests for SG.

SG flow-induced vibration testing is performed consistent with the requirements of the NuScale Comprehensive Vibration Assessment Program as described in the Comprehensive Vibration Assessment Program (CVAP) Technical Report, TR-0716-50439. Visual examination of the SG components is performed as specified in Table 5-1 of TR-0716-50439. This test is coordinated with Test #108. The CVAP is addressed in Section 3.9.2. The steam generators are discussed in Section 5.4.1.

System Function System Function Categorization Function Verified by Test #

None N/A N/A Prerequisites:

N/A Component Level Tests None Tier 2 14.2-164 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-73: Security Access Control Test # 73 Preoperational test is required to be performed once.

Security access control is described in NuScale Power, LLC, NuScale Design of Physical Security Systems, TR-0416-48929, Revision 0.

System Function System Function Categorization Function Verified by Test #

The security access controls support the security-related Component level test i.

security plan described in NuScale Design of Physical Security Systems, TR-0416-48929, Revision 0.

Prerequisites

i. Security access control boundary for the protected and vital areas, described in the security technical report, are established.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Verify an access control system with a Use authorized and unauthorized i. The access points do not allow numbered photo identification identification badges in all vital area access to unauthorized badges.

badge system which will control access points in the RXB and CRB ii. The access points allow authorized access to vital areas within the RXB identified in NuScale Design of Physical personnel.

and CRB to authorized personnel. Security Systems, TR-0416-48929, [ITAAC 03.16.04]

Revision 0.

System Level Tests None Tier 2 14.2-165 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-74: Security Detection and Alarm-Test # 74 Preoperational test is required to be performed once.

Security detection and alarm is described in NuScale Power, LLC, NuScale Design of Physical Security Systems, TR-0416-48929, Revision 0.

System Function System Function Categorization Function Verified by Test #

The security detection and alarm system security-related Component level test i-v acts to satisfy the functional requirements described in NuScale Design of Physical Security Systems, TR-0416-48929, Revision 0.

Prerequisites

i. Required security system site acceptance tests have been completed and approved.

Component Level Tests Test Objective Test Method Acceptance Criteria

i. Unoccupied vital areas will be Access to all unoccupied vital areas that i. Verify the access door is locked.

designed with locking devices and are identified in the NuScale Design of ii. Upon entry into the room verify an intrusion detection devices that Physical Security Systems, TR-0416- intrusion alarm is received in the annunciate in the central alarm 48929, Revision 0. central alarm station.

station. [ITAAC 03.16.05]

ii. Security alarm devices including a. Insert a signal real or simulated Verify alarm annunciation is received in transmission lines to annunciators tamper signal. the central alarm station for each test are tamper-indicating and self- b. Insert a signal real or simulated of a method. The alarm must indicate the checking. component failure for all alarm type and location of the alarm.

devices and transmission lines in the [ITAAC 03.16.07]

RXB and CRB.

c. Place all security alarm devices in the RXB and CRB on standby power.

iii. Intrusion detection and assessment Put all intrusion detection equipment Verify an audible and visual alarm is systems provides visual and audible described in NuScale Design of Physical received in the central alarm station.

alarm annunciation in the central Security Systems, TR-0416-48929, [ITAAC 03.16.08]

alarm station. Revision 0 into an alarm state.

iv. Intrusion detection system recording Place all intrusion detection equipment Verify the intrusion detection system equipment will record onsite security in the RXB and CRB in the following alarm recording system records each alarm to alarm annunciation including false conditions (as applicable to the include:

alarm, alarm check, and tamper equipment): a. Location of the alarm indication and the type of alarm, a. False alarm b. Type of alarm location, alarm circuit, date, and time.

b. Alarm check c. Alarm circuit

c. Tamper indication d. Date

e. Time (this test can be done in conjunction with audible and visual alarm testing)

[ITAAC 03.16.09]

v. Emergency exits in the RXB and CRB i. Attempt to enter each the RXB and i. Verify the locking device prevents will be alarmed with intrusion CRB exits. entry.

detection devices and secured by ii. Exit each of the RXB and CRB exits. ii. Verify the exit allows for prompt exit locking devices that allow prompt of the building and alarms in the egress during an emergency. central alarm station when opened.

[ITAAC 03.16.10]

System Level Tests None Tier 2 14.2-166 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-75: Initial Fuel Loading Precritical Test (Test #75)

Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading but prior to initial criticality.

Test Objectives

i. Identify the sequence for precritical testing (after fuel load and prior to criticality).

ii. The pre-critical tests are:

a. Reactor Coolant System Flow Measurement Test (Test #77)

b. NuScale Power Module Temperatures Test (Test #78)

c. Primary and Secondary System Chemistry Test (Test #79)

d. Control Rod Drive System - Manual Operation, Rod Speed, and Rod Position Indication Test (Test #80)

e. Control Rod Assembly (CRA) Drop Time Test (Test #81)

f. Pressurizer Spray Bypass Flow Test (Test #82)

Prerequisites None Test Method

i. Identify the specific plant conditions required for each precritical test procedure to maintain technical specification operability.

ii. Identify the prerequisites required for each precritical test procedure.

iii. Determine the test sequence for precritical testing based on technical specification requirements and test prerequisites.

Acceptance Criterion The sequence for precritical testing has been determined.

Tier 2 14.2-167 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-76: Initial Fuel Load Test (Test #76)

The Initial Fuel Load Test is required to be performed for each NPM.

This test is performed prior to initial fuel load.

Test Objectives

i. Conduct initial fuel load with no inadvertent criticality.

ii. Install fuel assemblies and control components at the locations specified by the design of the initial RXC.

Prerequisites

i. Plant systems required for initial fuel loading have completed preoperational testing.

ii. Plant systems required for initial fuel loading have been aligned per operations procedures.

iii. The design of the initial RXC that specifies the final core configuration of fuel assemblies and control components is completed.

iv. A core load sequence has been approved.

v. Neutron monitoring data from a previous NPM initial fuel loading or calculations showing the predicted response of monitoring channels are available for evaluating monitoring data.

vi. The lower RPV is installed in the RPV support stand.

vii. Control room communications are established.

viii. RXB radiation monitors are functional.

Test Method

i. The overall process of initial fuel loading will be supervised by a licensed senior reactor operator with no other concurrent duties.

ii. Install fuel and control components per approved procedures.

iii. Monitor boron concentration inside the RPV periodically during fuel load to ensure it satisfies TS.

iv. Monitor neutron counts during the load of each fuel assembly and plot an independent inverse count rate ratio for each source range detector after each fuel load assembly is loaded.

v. Verify neutron count data are consistent with calculations showing the predicted response. For fuel loading of the second NPM and all subsequent NPMs use data obtained from previous fuel loadings.

vi. Demonstrate the inverse count rate ratio does not show significant approach to criticality.

vii. Maintain the status of the core loading.

viii. Maintain communication between fuel handling personnel and the MCR.

Acceptance Criterion Each fuel assembly and control component is installed in the location specified by the design of the initial RXC.

Tier 2 14.2-168 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-77: Reactor Coolant System Flow Measurement Test (Test #77)

The Reactor Coolant System Flow Measurement test is required to be performed for each NPM.

This test is performed after initial fuel loading but prior to initial criticality.

Test Objective Verify that the RCS flow is sufficient to ensure adequate boron mixing in the RCS coolant.

Prerequisites

i. The core is installed.

ii. The NPM is fully assembled.

iii. The RCS is at hot zero power (RCS at normal operating pressure with RCS temperature at the maximum temperature obtainable when heated only by the MHS).

iv. The RCS flow meters have been calibrated.

Test Method Record RCS flow using MCR indication.

Acceptance Criterion The RCS flow at hot zero power (HZP) satisfies the minimum RCS flow assumed in the safety analysis.

Tier 2 14.2-169 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-78: NuScale Power Module Temperatures Test (Test #78)

Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading but prior to initial criticality.

Test Objectives

i. Perform a cross calibration of the RTDs monitored by the MPS listed in Table 7.1-2.

ii. Verify incore thermocouple resistance leakage satisfies manufacturers criteria.

Prerequisites

i. The core is installed.

ii. The NPM is fully assembled.

iii. The calibration of reactor coolant system RTDs has been completed.

Test Method

i. With the RCS at ambient temperature and isothermal conditions record the following data:

Main control room indication of RTD temperatures monitored by MPS

Main control room indication of incore thermocouples temperatures

Leakage resistance of the incore thermocouples ii. Increase RCS temperature by approximately 50°F.

iii. Record RTD and incore thermocouple data at isothermal conditions.

iv. Repeat data collection until RCS temperature is at the highest temperature obtainable using only the MHS.

v. Cross-calibrate RTD temperatures monitored by MPS that monitor the same variable.

Acceptance Criteria

i. The cross calibration of the reactor coolant system RTDs has been completed.

ii. The leakage resistance of the fixed incore detectors satisfies manufacturers recommendations.

Tier 2 14.2-170 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-79: Primary and Secondary System Chemistry Test (Test #79)

Startup test is required to be performed for each NPM.

This test is performed at approximately 25, 50, 75, and 100 percent reactor thermal power.

Test Objective Verify water quality in the primary system and secondary system using the PSS.

Prerequisites

i. The PSS instruments have been calibrated.

ii. The NPM is fully assembled.

iii. The RCS is at hot zero power (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).

Test Method

i. Use the PSS to sample the normal primary system sample points listed in Table 9.3.2-1.

ii. Use the PSS to sample the normal secondary system sample points listed in Table 9.3.2-3.

iii. To the extent practical, responses of PSS radiation monitors should be verified by laboratory analyses of grab samples taken at the same process location.

iv. Conduct the test at steady-state condition at approximately 25, 50, 75, and 100% reactor thermal power.

Acceptance Criterion The sample analysis satisfies the limits specified in plant procedures.

Tier 2 14.2-171 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-80: Control Rod Drive System - Manual Operation, Rod Speed, and Rod Position Indication Test (Test #80)

Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading but prior to initial criticality.

Test Objectives

i. Verify the ability to manually fully insert and fully withdraw individual control rod assemblies (CRAs) from the MCR.

ii. Verify CRA rod position indications provide indication of rod movement.

iii. Verify individual CRA position indications are within the required number of steps of their associated group position.

iv. Verify the rod insertion and withdrawal speeds are within design limits.

Prerequisites

i. The core is installed.

ii. The NPM is fully assembled.

iii. The RCS is at hot zero power (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).

iv. All RCS temperatures satisfy the minimum technical specification temperature for criticality.

v. The nuclear instrumentation system is calibrated and operable.

vi. The SDM is within the limits specified in the core operating limits report.

Test Method

i. Individually withdraw and insert each shutdown bank and regulating bank from the MCR a sufficient number of steps to verify that the individual CRA positions are within the required number of steps of their group position as required by TS.

Only the tested bank will be withdrawn. All other banks are fully inserted. Repeat the test until all shutdown banks and regulating banks are tested.

ii. With all shutdown and regulating banks fully inserted, fully withdraw and then fully insert one CRA. Repeat these steps until all CRAs are tested.

Acceptance Criteria

i. All CRAs can be individually fully withdrawn and fully inserted from the MCR.

ii. Individual CRA positions agree with the control rod demand position within design limits for the full range of CRA travel.

iii. Individual CRA position indications are within the number of steps of their associated group position as required by TS.

iv. The CRA insertion and withdrawal speeds are within design limits.

Tier 2 14.2-172 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-81: Control Rod Assembly Drop Time Test (Test #81)

Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading but prior to initial criticality.

Test Objective Verify each CRA satisfies the CRA drop time acceptance criteria for RCS flow at 0% reactor thermal power.

Prerequisites

i. The core is installed.

ii. The NPM is fully assembled.

iii. The RCS is at hot zero power (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).

iv. All RCS temperatures satisfy the minimum technical specification temperature for criticality.

v. The nuclear instrumentation system is calibrated and operable.

vi. The SDM is within the limits specified in the core operating limits report.

vii. A CRA drop time acceptance criteria for 0% thermal reactor power has been developed and is in agreement with the technical specification CRA drop time surveillance requirement.

Test Method

i. Withdraw each individual CRA.

ii. Interrupt the electrical power to the associated CRDM.

iii. Measure the CRA drop time.

Acceptance Criteria

i. Each CRA drop time is less than or equal to the CRA drop time acceptance criteria for HZP.

ii. The arithmetic average of all CRA drop times is within TS limits.

Tier 2 14.2-173 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-82: Pressurizer Spray Bypass Flow Test (Test #82)

Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading but prior to initial criticality.

Test Objective Verify the pressurizer spray bypass flow rate is adequate to prevent thermal fatigue of the spray line components and provide sufficient mixing in the pressurizer to maintain pressurizer water chemistry similar to the rest of the RCS while avoiding unnecessary energization of the pressurizer heaters.

Prerequisites

i. The core is installed.

ii. The NPM is fully assembled.

iii. The RCS is at hot zero power (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).

Test Method

i. With the automatic pressurizer spray valve closed, adjust the manual spray bypass valve to maintain a continuous spray bypass flow of approximately one gpm.

ii. If the continuous bypass spray flow requires the operation of the pressurizer backup heaters to maintain the pressurizer pressure setpoint, throttle close the bypass valve until pressurizer pressure is maintained by the proportional heaters.

Acceptance Criterion The spray bypass valve is throttled to maintain the required bypass flow.

Tier 2 14.2-174 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-83: Initial Criticality Test (Test #83)

Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading.

Test Objective Achieve initial criticality in a controlled manner.

Prerequisites

i. The RCS is at hot zero power (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).

ii. All RCS temperatures satisfy the minimum technical specification temperature for criticality.

iii. The nuclear instrumentation system is calibrated and operable.

iv. The SDM is within the limits specified in the core operating limits report.

v. An estimated critical position (calculation) has been performed.

vi. RCS measured boron is at or near the desired estimated critical position value.

vii. The shutdown banks and the regulating banks are fully inserted.

Test Method

i. Shutdown banks are withdrawn in sequence using the sequence of a normal plant startup. Gather data to plot the inverse count rate ratio. The inverse count rate ratio is used to monitor reactivity.

ii. Once all shutdown banks are fully withdrawn, then the regulating bank is withdrawn using the sequence of a normal plant startup. The inverse count rate ratio is plotted to monitoring reactivity for the approach to criticality.

iii. After criticality is obtained, the regulating bank is confirmed to be above the TS regulating group insertion limit.

iv. Should criticality be reached with the regulating bank below the insertion limit specified by the core operating limits requirement, the limiting condition of operation test exception is invoked. The RCS boron will be increased until the regulating bank is withdrawn sufficiently to meet the insertion limit.

Acceptance Criterion The reactor is critical with the regulating banks above their technical specification insertion limit.

Tier 2 14.2-175 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-84: Post - Critical Reactivity Computer Checkout Test (Test #84)

Startup test is required to be performed for each NPM.

This test is performed after initial criticality.

Test Objective Verify proper operation of the reactivity computer to measure reactivity changes in the core during low-power testing.

Prerequisites

i. The reactor is critical with the neutron flux level within the range for low-power physics testing.

ii. The RCS temperature and pressure are stable at the normal no-load values.

iii. The neutron flux level and RCS boron concentration are stable.

iv. The reactivity computer is installed and internal reactivity computer checks have been completed.

Test Method

i. Withdraw the regulating bank to achieve a positive startup rate below TS limits.

ii. Measure the reactor period or doubling time.

iii. Reinsert the regulating bank to re-establish the initial steady-state neutron flux.

iv. Measure the negative reactor period or halving time.

v. Validate the core response against the reactivity computer input delayed neutron fractions and prompt neutron lifetime using pre-determined test criteria.

vi. Adjust and recalibrate reactivity computer until acceptance criteria is met.

Acceptance Criterion The reactivity computer is calibrated.

Tier 2 14.2-176 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-85: Low Power Test Sequence (Test #85)

Startup test is required to be performed for each NPM.

This test is performed before initial criticality.

Test Objectives

i. Identify the sequence for low-power testing.

ii. The low-power tests are:

a. Determination of Zero-Power Physics Testing Range Test (Test #86)

b. All Rods Out (ARO) Boron Endpoint Determination Test (Test #87)

c. Isothermal Temperature Coefficient Measurement Test (Test #88)

d. Bank Worth Measurement Test (Test #89)

Prerequisites None Test Method For each of the tests identified in the test objectives above:

i. Identify the specific plant conditions required for each low-power test procedure to maintain technical specification operability.

ii. Identify the prerequisites required for each low-power test procedure.

iii. Determine the test sequence for low-power testing based on technical specification requirements and test prerequisites.

Acceptance Criterion The sequence for low-power testing has been determined.

Tier 2 14.2-177 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-86: Determination of Zero-Power Physics Testing Range Test (Test #86)

Startup test is required to be performed for each NPM.

This test is performed after initial criticality.

Test Objectives

i. Determine the reactor flux level at which the point of nuclear heating is detectable.

ii. Establish the range of neutron flux in which HZP reactivity measurements are to be performed.

Prerequisites

i. The reactor is critical with the neutron flux level at steady-state below the expected level of nuclear heating.

ii. The RCS temperature and pressure is steady-state at the normal HZP conditions.

iii. The RCS boron concentration is steady-state.

iv. The reactivity computer is operational and recording the core average neutron flux level.

v. The regulating bank is positioned to allow reactivity changes by rod motion alone.

Test Method

i. Withdraw the regulating bank to establish a slow startup rate allowing neutron flux level to increase until nuclear heating is observed.

ii. Record the reactivity computer neutron flux level and the corresponding MCR flux indication at which nuclear heating occurs.

iii. Insert the regulating bank to establish a reactivity computer flux level about one-third of the value at which nuclear heating was observed. This flux level becomes the maximum value for the zero-power testing range.

Acceptance Criterion The zero power testing range flux level is determined.

Tier 2 14.2-178 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-87: All Rods Out Boron Endpoint Determination Test (Test #87)

Startup test is required to be performed for each NPM.

This test is performed after initial criticality.

Test Objective Determine the critical RCS boron concentration for ARO (fully withdrawn shutdown banks and regulating banks) at HZP.

Prerequisites

i. The reactor is critical with the neutron flux level at steady-state below the expected level of nuclear heating.

ii. The RCS temperature and pressure is steady-state at the normal HZP conditions.

iii. The RCS boron concentration is steady-state.

iv. The reactivity computer is operational and recording the core average neutron flux level.

Test Method

i. Add a pre-determined volume of borated water to the RCS and withdraw the regulating bank to maintain critical conditions. The final regulating bank position will be near fully withdrawn and will limit the usable positive reactivity remaining in the rods with the reactor critical.

ii. Measure the just-critical boron concentration by chemical analysis.

iii. Fully withdraw the regulating bank without adjusting the boron concentration. Measure and calculate the change in reactivity for ARO and the RCS temperature difference from program TAVG, due to an equivalent change in boron concentration. Add the equivalent boron change to the just-critical boron concentration to yield the endpoint for ARO.

Acceptance Criterion The measured value for the ARO boron endpoint satisfies the design value contained within the test acceptance criteria.

Tier 2 14.2-179 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-88: Isothermal Temperature Coefficient Measurement Test (Test #88)

Startup test is required to be performed for each NPM.

This test is performed after initial criticality.

Test Objectives

i. Determine the isothermal temperature coefficient.

ii. Calculate the moderator temperature coefficient.

Prerequisites

i. The reactor is critical with the neutron flux level at steady-state below the expected level of nuclear heating.

ii. The RCS temperature and pressure is steady-state at the normal HZP conditions.

iii. The RCS boron concentration is steady-state.

iv. The reactivity computer is operational and recording the core average neutron flux level.

v. The regulating rod bank is positioned near fully withdrawn (near their ARO position).

Test Method

i. Vary RCS temperature (heatup/cooldown) while maintaining rods and boron concentration constant.

ii. Monitor reactivity results and determine the isothermal temperature coefficient.

iii. Calculate the moderator temperature coefficient using the isothermal temperature coefficient and design values.

Acceptance Criterion The moderator temperature coefficient is within the limits specified in the core operating limits report.

Tier 2 14.2-180 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-89: Bank Worth Measurement Test (Test #89)

Startup test is required to be performed for each NPM.

This test is performed after initial criticality.

Test Objectives

i. Measure the integral and differential worth of the reference bank (the test bank with the highest predicted worth).

ii. Measure the worth of the remaining shutdown and regulating banks by control rod exchange (rod swap).

Prerequisites

i. The reactor is critical with the neutron flux level at steady-state within the range for HZP physics testing.

ii. The RCS temperature and pressure is steady-state at the normal HZP conditions.

iii. The RCS boron concentration is steady-state.

iv. The reactivity computer is operational and recording the core average neutron flux level.

v. The regulating rod banks are positioned near fully withdrawn (near their ARO position).

Test Method

i. The referenced bank rod worth measurement is made by performing a slow controlled boron dilution while the reference bank is inserted to maintain criticality. The rod worth is measured using the reactivity computer. During boron dilution the reference bank step insertions maintain neutron flux within the zero-power physics test range until the referenced bank is fully inserted.

ii. A test bank rod worth measurement is made by inserting the test bank while the reference bank is withdrawn. The test bank worth is determined by the final position of the referenced bank.

Acceptance Criterion The measured worth for each individual bank, and sum of bank worths, is consistent with the predicted value within the test acceptance criteria.

Tier 2 14.2-181 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-90: Power-Ascension Test (Test #90)

Startup test is required to be performed for each NPM.

This test is performed prior to power-ascension testing.

Test Objective Identify the sequence for the following power-ascension tests.

a. Core Power Distribution Map Test (Test #91)

b. Neutron Monitoring System Power Range Flux Calibration Test (Test #92)

c. Reactor Coolant System Temperature Instrument Calibration Test (Test #93)

d. Reactor Coolant System Flow Calibration Test (Test# 94)

e. Radiation Shield Survey Test (Test #95)

f. Reactor Building Ventilation System Capability Test (Test #96)

g. Thermal Expansion Test (Test #97)

h. Control Rod Assembly Misalignment Test (Test #98)

i. Steam Generator Level Control System Test (Test #99)

j. Ramp Change in Load Demand Test (Test #100)

k. Step Change in Load Demand Test (Test #101)

l. Loss of Feedwater Heater Test (Test #102)

m. 100 Percent Load Rejection Test (Test #103)

n. Reactor Trip from 100 Percent Power Test (Test #104)

o. Island Mode Test for NuScale Power Module #1 (Test #105)

p. Island Mode Test for Multiple NuScale Power Modules (Test #106)

q. Remote Shutdown Workstation Test (Test #107)

r. Reactor Module Vibration Test (Test #108)

Prerequisites None Test Method

i. Identify the specific plant conditions required for each power-ascension test procedure to maintain technical specification operability.

ii. Identify the prerequisites required for each power-ascension test procedure.

iii. Determine the test sequence for power-ascension testing based on technical specification requirements and test prerequisites.

Acceptance Criterion The sequence for power-ascension testing has been determined.

Tier 2 14.2-182 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-91: Core Power Distribution Map Test (Test #91)

Startup test is required to be performed for each NPM.

This test is performed at approximately 25, 50, 75, and 100 percent reactor thermal power Test Objectives

i. Obtain a core power distribution map during power ascension.

ii. Using the data from the core power distribution map verify core power distribution is consistent with design predictions and associated technical specifications limits.

Prerequisites

i. The ICIS is operational.

ii. The NPM is operating in a steady-state condition at the specified power level.

iii. Maintain reactor power, TAVG, and pressurizer level constant during data collection.

Test Method

i. With the plant at power levels of approximately 25, 50, 75, and 100 percent of reactor thermal power, obtain a core power distribution map during power ascension using the MCS and instrument input from the in-core self-powered neutron detectors.

ii. Use data from the in-core maps to verify that core power distribution is consistent with design predictions and technical specifications limits.

Acceptance Criterion Core power distribution is consistent with design predictions and technical specifications limits.

Tier 2 14.2-183 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-92: Neutron Monitoring System Power Range Flux Calibration Test (Test #92)

Startup test is required to be performed for each NPM.

This test is performed at approximately 25, 50, 75 and 100 percent reactor thermal power.

Test Objective Calibrate the NMS power range neutron flux signals during power ascension.

Prerequisites

i. The ICIS is operational.

ii. The NPM is operating in a steady-state condition at the specified power level.

Test Method

i. With the plant at power levels of approximately 25, 50, 75 and 100 percent of reactor thermal power, record the following data:

power range neutron flux from the ICIS self-powered neutron detectors

NMS power range (linear power) signal ii. Maintain reactor power, TAVG, and pressurizer level constant during data collection.

iii. Calibrate the NMS neutron flux power range (linear power) signal using the recorded data.

Acceptance Criterion The NMS neutron flux power range (linear power) signal has been calibrated.

Tier 2 14.2-184 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-93: Reactor Coolant System Temperature Instrument Calibration Test (Test #93)

Startup test is required to be performed for each NPM.

This test is performed at approximately 25, 50, 75, and 100 percent reactor thermal power.

Test Objective Calibrate narrow range RCS hot leg temperature instruments, wide range RCS hot leg temperature instruments, and narrow range RCS cold leg temperature instruments.

Prerequisites

i. The ICIS is operational.

ii. The NPM is operating in a steady-state condition at the specified power level.

Test Method

i. With the plant at power levels of approximately 25, 50, 75, and 100 percent of reactor thermal power, record the following data:

NMS flux power range (linear power) signal

RCS narrow range hot leg temperature

RCS wide range hot leg temperature

RCS narrow range cold leg temperature

ICIS core inlet and outlet temperature ii. Maintain reactor power, TAVG, and pressurizer level at steady-state during data collection.

iii. Calibrate the RCS narrow range and wide range hot leg temperature instruments and the RCS narrow range cold leg temperature using the recorded data.

Acceptance Criterion The RCS hot and cold let temperature instruments have been calibrated.

Tier 2 14.2-185 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-94: Reactor Coolant System Flow Calibration Test (Test #94)

Startup test is required to be performed for each NPM.

This test is performed at approximately 25, 50, 75, and 100 percent reactor thermal power.

Test Objective Calibrate the RCS flow instruments during power ascension.

Prerequisites

i. The ICIS is operational.

ii. The NPM is operating in a steady-state condition at the specified power level.

Test Method

i. With the plant at power levels of approximately 25, 50, 75, and 100 percent of reactor thermal power, record the following data:

NMS flux power range (linear power) signal

RCS narrow range hot leg temperature

RCS narrow range cold leg temperature

ICIS core inlet and outlet temperature ii. Maintain reactor power, TAVG, and pressurizer level at steady state during data collection.

iii. Calibrate the RCS flow instruments using the recorded data.

Acceptance Criterion The RCS flow instruments have been calibrated.

Tier 2 14.2-186 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-95: Radiation Shield Survey Test (Test #95)

Startup test is required to be performed for each NPM.

This test is performed at approximately 25, 50, and 100 percent reactor thermal power.

Test Objective Verify the adequacy of concrete radiation shields in the RXB designed to protect personnel from radiation originating from sources within the reactor vessel.

Prerequisites

i. Radiation survey instruments are calibrated.

ii. The NPM is operating in a steady-state condition at the specified power level.

Test Method

i. Measure gamma and neutron radiation dose rates at designated locations at approximately 25, 50, and 100 percent reactor thermal power.

ii. The designated locations are the accessible areas outside permanent concrete radiation shields in the RXB. (reference Figure 12.3-1a for the RXB radiation zone map)

Acceptance Criterion Radiation dose rates are consistent with design expectations.

Tier 2 14.2-187 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-96: Reactor Building Ventilation System Capability (Test #96)

Startup test is required to be performed for each NPM.

This test is performed at approximately 50 and 100 percent reactor thermal power.

Test Objective Verify that the RXB ventilation system maintains the design environment in areas containing equipment that is environmentally qualified for a harsh or mild environment.

Prerequisite The NPM is operating in a steady-state condition at the specified power level.

Test Method

i. With the plant at power levels of approximately 50 and 100% of reactor thermal power and RXB ventilation system in normal lineup, record temperature and humidity for the environmental qualification zones listed in Table 3.11-2 that are not under the bioshield.

ii. With the plant at power levels of approximately 50 and 100% of reactor thermal power and RXB ventilation system in normal lineup, record the temperature and humidity in the rooms containing electrical equipment qualified for a mild environment.

Acceptance Criteria

i. Room temperature and humidity in environmental qualification zones listed in Table 3.11-2 that are not under the bioshield satisfy the indoor design conditions for the RXB ventilation system contained in Table 9.4.2-2.

ii. Room temperature and humidity in rooms containing electrical equipment qualified for a mild environment satisfy the indoor design conditions for the RXB ventilation system contained in Table 9.4.2-2.

Tier 2 14.2-188 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-97: Thermal Expansion Test (Test #97)

Startup test is required to be performed for each NPM.

This test is performed during plant heatup and cooldown.

Test Objectives

i. Verify that ASME Code Class 1, 2, and 3 system piping can expand without obstruction and that expansion is within design limits. All ASME Code Class 1, 2, and 3 system piping is within the RXB.

ii. Verify that high-energy piping inside the RXB can expand without obstruction and that expansion is within design limits.

Prerequisite Temporary instrumentation is installed on piping outside the NPM as required to monitor the deflections for the piping under test.

Test Method

i. Thermal expansion testing is performed in accordance with ASME OM Standard, Part 7 as discussed in Section 3.9.2.1.2.

ii. Record deflection data during plant heatup and cooldown.

iii. Identify support movements by recording hot and cold positions of the supports.

iv. All tested piping is within the RXB.

v. All tested piping is outside the NPM.

vi. All tested piping is contained within the MSS, FWS, ABS, PSS, and CVCS.

Acceptance Criteria For the piping systems tested:

i. There is no evidence of blocking of the thermal expansion of piping or component, other than by installed supports, restraints, and hangers.

ii. Spring hanger movements must remain within the hot and cold setpoints and supports must not become fully retracted or extended.

iii. Piping and components return to their approximate baseline cold position.

Tier 2 14.2-189 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-98: Control Rod Assembly Misalignment (Test #98)

Startup test is required to be performed for each NPM.

This test is performed between 30 and 50 percent reactor thermal power.

Test Objectives

i. Verify that core thermal and nuclear parameters at 50% reactor thermal power are in accordance with predictions with a single high-worth rod fully inserted, during rod movement, and following return of the rod to its bank position.

ii. Verify the capability of the in-core neutron flux instrumentation to detect a control rod misalignment equal to or less than the technical specification limits at 50% and 100% reactor thermal power.

iii. Monitor the power distribution following the recovery of a misaligned CRA.

Prerequisites

i. The reactor is operating at approximately 50% reactor thermal power and has been at that power for a sufficient time to reach xenon equilibrium.

ii. The reactor power level, reactor coolant system boron concentration, and temperature are stable.

iii. The regulating and shutdown banks are positioned as required for the specific measurement, near fully withdrawn for CRA insertion, and at their respective insertion limits for CRA withdrawal.

Test Method

i. For the CRA insertion, insert a group of selected CRAs, one at a time, first to the limit of misalignment specified in TS, then fully inserted, and finally restored to the bank position. Compensate for reactivity changes by dilution and boration as required.

ii. For the CRA withdrawal, withdraw one or more selected CRAs, one at a time, to the fully withdrawn position. Compensate for reactivity changes by boration and dilution as required.

iii. Record incore and excore instrumentation signals to determine their response and to determine the power distribution and power peaking factors prior to control rod assembly misalignment, at partial misalignment, at full misalignment, and periodically after restoration to normal.

Acceptance Criteria Measured power distributions and power peaking factors are within technical specification limits and are consistent with the predictions.

Tier 2 14.2-190 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-99: Steam Generator Level Control Test (Test #99)

Startup test is required to be performed for each NPM.

This test is performed at approximately 25, 50, 75, and 100 percent reactor thermal power.

Test Objective Verify the stability of the automatic SG level control system by introducing simulated transients at various power levels during the ascension to full power.

Prerequisite The NPM is operating in a steady-state condition at the specified power level.

Test Method

i. Simulate an SG level transient by changing the level setpoint at approximately 25, 50, 75, and 100% reactor thermal power.

ii. Record the steam generator level control response when the control system is returned to automatic control.

iii. Adjustments to the control systems are made, if necessary, prior to proceeding to the next power plateau.

Acceptance Criteria

i. During recovery from a simulated steam generator level transient, SG level control response is consistent with the design for the following:

a. overshoot or undershoot to the new level.

b. time required to achieve the new level.

c. error between the actual level and control setpoint.

d. feedwater pump discharge pressure oscillations.

ii. Water hammer indications:

a. Audible indications of water hammer are not observed.

b. No damage to pipe supports or restraints.

c. No damage to equipment.

d. No equipment leakage as a result of the steam generator level transient.

Tier 2 14.2-191 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-100: Ramp Change in Load Demand (Test #100)

Startup test is required to be performed for each NPM.

This test is performed at approximately 25, 50, 75, and 100 percent reactor thermal power.

Test Objectives

i. Verify the ability of the plant automatic control systems to sustain a ramp increase in load demand.

ii. Assess the dynamic response of the plant for ramp increase in load demand.

Prerequisites

i. The NPM is operating in a steady-state condition at the designated power level.

ii. The plants electrical distribution system is aligned for normal operation.

iii. Reactor, turbine, and secondary control systems are in automatic mode.

Test Method

i. Use the main control room turbine controls to provide a 5% of full power per minute load increase in demand at approximately 25, 50, and 75% reactor thermal power.

ii. Use the main control room turbine controls to provide a 5% of full power per minute load decrease in demand at approximately 25, 50, and 75, and 100% reactor thermal power.

Acceptance Criteria

i. The turbine does not trip.

ii. The reactor does not trip.

iii. The main steam safety valves do not open.

iv. The turbine does not overspeed.

v. The primary and secondary control systems, with no manual intervention, maintain reactor power, reactor coolant system temperatures, pressurizer pressure and level, and SG levels and pressures within acceptable ranges during and following the transient.

vi. Control system response is reviewed and compared to expected performance. Necessary adjustments to the control systems have been made prior to proceeding to the next power plateau.

vii. Water hammer indications

a. Audible indications of water hammer are not observed.

b. No damage to pipe supports or restraints.

c. No damage to equipment.

d. No equipment leakage as a result of the ramp change.

Tier 2 14.2-192 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-101: Step Change in Load Demand Test (Test #101)

Startup test is required to be performed for each NPM.

This test is performed at approximately 25, 50, 75, and 100 percent reactor thermal power.

Test Objectives

i. Verify the ability of the plant automatic control systems to sustain step load increases and step load decreases in demand.

ii. Assess the dynamic response of the plant for a load step demand.

Prerequisites

i. The NPM is operating in a steady-state condition at the specified power level.

ii. The plants electrical distribution system is aligned for normal operation.

iii. Reactor, turbine, and secondary control systems are in automatic mode.

Test Method

i. Use the MCR turbine controls to provide a 10% step load increase in demand at approximately 25, 50, and 75% reactor thermal power.

ii. Use the MCR turbine controls to provide a 10% step load decrease in demand at approximately 25, 50, 75, and 100%

reactor thermal power.

Acceptance Criteria

i. The turbine does not trip.

ii. The reactor does not trip.

iii. The main steam safety valves do not open.

iv. The turbine does not overspeed.

v. The primary and secondary control systems, with no manual intervention, maintain reactor power, RCS temperatures, pressurizer pressure and level, and SG levels and pressures within acceptable ranges during and following the transient.

vi. Control system response is reviewed and compared to expected performance. Necessary adjustments to the control systems have been made prior to proceeding to the next power plateau.

vii. Water hammer indications

a. Audible indications of water hammer are not observed.

b. No damage to pipe supports or restraints.

c. No damage to equipment.

d. No equipment leakage as a result of the step load change.

Tier 2 14.2-193 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-102: Loss of Feedwater Heater Test (Test #102)

Startup test is required to be performed for each NPM.

This test is performed at approximately 50 and 90 percent reactor thermal power.

Test Objectives

i. Verify the ability of the plant automatic control systems to sustain a loss of the high pressure feedwater heater during power operation.

ii. Assess the dynamic response of the plant for the loss of the high pressure feedwater heater.

Prerequisites

i. The NPM is operating in a steady-state condition at the specified power level.

ii. The plants electrical distribution system is aligned for normal operation.

iii. Reactor, turbine, and secondary control systems are in automatic mode.

Test Method Close the turbine generator extraction steam supply isolation valve to the high pressure feedwater heater from the main control room at approximately 50 and 90% reactor thermal power.

Acceptance Criteria

i. The reactor does not trip.

ii. The turbine does not trip.

iii. The main steam safety valves do not open.

Tier 2 14.2-194 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-103: 100 Percent Load Rejection Test (Test #103)

Startup test is required to be performed for each NPM.

This test is performed at approximately 100 percent reactor thermal power.

Test Objectives

i. Verify the ability of the plant automatic control systems to sustain a 100% load rejection from full power.

ii. Assess the dynamic response of the plant for a 100% power load rejection.

Prerequisites

i. The NPM is operating in a steady-state condition at full reactor thermal power.

ii. The plants electrical distribution system is aligned for normal operation.

iii. Reactor, turbine, and secondary control systems are in automatic mode.

Test Method Manually trip the generator output breaker to provide a 100 percent load rejection.

Acceptance Criteria

i. The turbine trips.

ii. The reactor does not trip.

iii. The main steam safety valves do not open.

iv. The turbine does not overspeed beyond design limits.

v. The turbine generator bypass valve opens and modulates steam flow to the condenser to maintain steam generator pressure.

vi. The FWS automatically provides the necessary feedwater flow to the steam generator.

vii. Water hammer indications

a. Audible indications of water hammer are not observed

b. No damage to pipe supports or restraints

c. No damage to equipment

d. No equipment leakage as a result of the load rejection.

Tier 2 14.2-195 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-104: Reactor Trip from 100 Percent Power Test (Test #104)

Startup test is required to be performed for each NPM.

This test is performed at 100 percent reactor thermal power.

Test Objectives

i. Verify the ability of the NPM to sustain a reactor trip from 100% reactor thermal power and automatically cool the RCS to mode 3 (all RCS temperatures < 420 °F).

ii. Assess the dynamic response of the plant to the reactor trip.

Prerequisites

i. The NPM is operating in a steady-state condition at full reactor thermal power.

ii. The plants electrical distribution system is aligned for normal operation.

Test Method

i. Manually trip the reactor from the MCR.

ii. Allow the RCS to cool to mode 3.

Acceptance Criterion

i. The reactor trips.

ii. The CIVs close.

iii. The decay heat removal valves open.

iv. The turbine generator bypass valve operates to prevent opening of the main steam safety valve.

v. The turbine speed does not exceed overspeed design limits.

vi. The reactor vent valves do not open.

vii. Water hammer indications

a. Audible indications of water hammer are not observed

b. No damage to pipe supports or restraints

c. No damage to equipment

d. No equipment leakage as a result of the reactor trip viii. The RCS cools to a stable condition in mode 3 without operator intervention.

Tier 2 14.2-196 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-105: Island Mode Test for NuScale Power Module #1(Test #105)

This startup test is required to be performed for the first NPM in power operation. No other NPMs are in power operation. Test #105 is performed once per facility. Startup Test #106 tests island mode for multiple NPMs.

This test is performed at 100 percent reactor thermal power. Island mode operation is described in Section 8.3.1.1.1 Test Objective for the first NPM in power operation

i. Verify the first NPM in power operation can operate independently from an offsite transmission grid after transition from the transmission grid to island mode.

ii. Verify plant electrical loads may be transitioned from island mode to an offsite transmission grid without interruption to the operation of the first NPM in power operation.

Prerequisites The first NPM in power operation is in normal operation at 100 percent reactor thermal power.

Test Method Simulate a loss of the transmission grid by opening the switchyard supply breakers (reference Figures 8.3-2a and 8.3-2b).

Acceptance Criterion i a. The turbine generator associated with the NPM under test does not trip and changes from droop mode control to isochronous mode to control the loads on site.

b. The first NPM in power operation remains at approximately 100 percent reactor thermal power using turbine generator bypass operation.

c. Electrical power to plant loads is uninterrupted without loss of voltage or automatic bus transfers.

d. The auxiliary AC power source starts automatically but does not automatically load its associated bus.

ii. The plant electrical loads are transitioned back to the external offsite grid connection when it becomes available.

Tier 2 14.2-197 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-106: Island Mode Test for Multiple NuScale Power Modules (Test #106)

This startup test is required to be performed once with multiple (at least two) NPMs in operation. Test #106 is performed once per facility. Startup Test #105 tests island mode for a single NPM.

COL Item 14.2-7: A COL applicant that references the NuScale Power Plant design certification will select the plant configuration to perform the Island Mode Test (number of NuScale Power Modules in service).

This test is performed at 100 percent reactor thermal power for all NPMs under test. Island mode operation is described in Section 8.3.1.1.1 Test Objective for multiple NPM in operation:

i. Verify all NPMs under test can operate independently from an offsite transmission grid after transition from the transmission grid to island mode.

ii. Verify plant electrical loads may be transitioned from island mode to an offsite transmission grid without interruption to the operation of the service unit NPM.

Prerequisites The NPMs selected for test are in normal operation at 100 percent reactor thermal power.

Test Method Simulate a loss of the transmission grid by opening the switchyard supply breakers (reference Figures 8.3-2a and 8.3-2b).

Acceptance Criterion

i. a. The service unit turbine generator transitions to island mode by changing from droop mode control to isochronous mode control to control the load on the 13.8kV bus it is supplying.

b. The service unit NPM remains at approximately 100 percent reactor thermal power using turbine generator bypass operation.

c. The non-service unit turbine generators trip.

d. The non-service unit NPMs power reduces to approximately 95% percent reactor thermal power using turbine generator bypass operation

e. Electrical power to plant loads is uninterrupted without loss of voltage or automatic bus transfers.

d. The auxiliary AC power source starts automatically but does not automatically load its associated bus.

ii. The plant electrical loads are successfully transitioned back to an external offsite grid connection when it becomes available Tier 2 14.2-198 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-107: Remote Shutdown Workstation Test (Test #107)

Startup test is required to be performed for each NPM.

This test is performed at approximately 10 - 20 percent reactor thermal power.

Test Objectives

i. Verify the NPM safety-related controls can be disabled at the remote shutdown station.

ii. Verify the NPM nonsafety-related controls are functional at the remote shutdown station.

Prerequisites

i. Communication exists between the MCR and the remote shutdown station.

ii. The reactor is operating in a steady-state condition at 10 - 20% reactor thermal power.

Test Method

i. Using the appropriate operating procedure, the operator manually trips the reactor under test before leaving the MCR.

ii. Using the appropriate operating procedure, the operator uses manual switches in the remote shutdown station to isolate the module protection system manual actuation switches, override switches, and the enable nonsafety control switches for each nuclear power modules module protection system in the MCR to prevent spurious actuation of equipment due to fire damage.

Acceptance Criteria

i. An operator verifies that the module protection switch controls in the MCR have been disabled.

The displays in the remote shutdown station verify the following NPM status:

ii. The reactor is tripped.

iii. All CIVs are closed.

iv. The DHRS actuation valves are open.

v. All RCS temperatures cool to less than 420°F (mode 3, safe shutdown) without operator action.

vi. Safety-related components cannot be operated from the remote shutdown station.

vii. The nonsafety-related controls in the remote shutdown station controls can be used to place the plant in a configuration specified by the appropriate operating procedure.

Tier 2 14.2-199 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-108: NuScale Power Module Vibration Test (Test #108)

This startup test is required to be performed once for NPM #1. This test supports FOAK testing described in Section 14.2.3.3.

This test is performed at 100 percent reactor thermal power. NuScale Power Module vibration testing is described in Sections 3.9.2.1.1.1, 3.9.2.3 and 3.9.2.4. and Reference 3.9-5 NuScale Power, LLC, Comprehensive Vibration Assessment Program (CVAP) Technical Report, TR-0716-50439.

Test Objective for NPM #1

i. Perform vibration testing of DHRS steam piping at 100 percent reactor thermal power as described in TR-0716-50439, Section 4.3, to verify vibration amplitudes in the DHRS steam piping confirm the acoustic resonance analysis results described in TR-0716-50439 Section 4.3.

ii. Perform visual examination of the NuScale Power Module components specified in Table 5-1 of TR-0716-50439.

Prerequisites

i. The DHRS steam piping is instrumented to obtain acoustic resonance (AR) data.

Test Method

i. Operate the NuScale Power Module for a sufficient duration at 100 percent power to ensure one million vibration cycles for the component with the lowest structural natural frequency.

ii. Monitor the vibration of the DHRS steam piping. If an unacceptable vibration response develops any time during initial startup testing, the test conditions will be adjusted to stop the vibration and the reason for the vibration anomaly will be investigated prior to continuing with the testing.

iii. Disassemble the NuScale Power Module and performed a visual examination of the module components specified in Table 5-1 of TR-0716-50439.

Acceptance Criterion

i. Measured vibration amplitudes in the DHRS steam piping confirm the acoustic resonance analysis results described in TR-0716-50439 Section 4.3.

ii. Visual examination results of module components satisfy the acceptance criteria of Table 5-1 of TR-0716-50439.

Tier 2 14.2-200 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-109: List of Test Abstracts Test Number System Abbreviation Test Abstract 1 SFPCS Spent Fuel Pool Cooling System 2 PCUS Pool Cleanup System 3 RPCS Reactor Pool Cooling System 4 PSCS Pool Surge Control System 5 UHS Ultimate Heat Sink 6 PLDS Pool Leakage Detection System 7 RCCWS Reactor Component Cooling Water System 8 CHW Chilled Water System 9 ABS Auxiliary Boiler System 10 CWS Circulating Water System 11 SCW Site Cooling Water System 12 PWS Potable Water System 13 UWS Utility Water System 14 DWS Demineralized Water System 15 NDS Nitrogen Distribution System 16 SAS Service Air System 17 IAS Instrument Air System 18 CRHS Control Room Habitability System 19 CRVS Normal Control Room HVAC System 20 RBVS Reactor Building HVAC System 21 RWBVS Radioactive Waste Building HVAC System 22 TBVS Turbine Building Ventilation 23 RWDS Radioactive Waste Drain System 24 BPDS Balance-of-Plant Drains 25 FPS Fire Protection System 26 FDS Fire Detection 27 MSS Main Steam 28 CFWS Feedwater System 29 FWTS Feedwater Treatment 30 CPS Condensate Polisher Resin Regeneration System 31 HVD Heater Vents and Drains 32 CARS Condenser Air Removal System 33 TGS Turbine Generator 34 TLOS Turbine Lube Oil System 35 LRWS Liquid Radioactive Waste System 36 GRWS Gaseous Radioactive Waste System 37 SRWS Solid Radioactive Waste System 38 CVCS Chemical and Volume Control System 39 BAS Boron Addition System 40 MHS Module Heatup System 41 CES Containment Evacuation System 42 CFDS Containment Flooding and Drain System 43 CNTS Containment System 44 CRDS Control Rod Drive System Flow-Induced Vibration 45 RVI Reactor Vessel Internals Flow-Induced Vibration 46 RCS Reactor Coolant System 47 ECCS Emergency Core Cooling System 48 DHRS Decay Heat Removal System 49 ICIS In-core Instrumentation 50 MAE Module Assembly Equipment Tier 2 14.2-201 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-109: List of Test Abstracts (Continued)

Test Number System Abbreviation Test Abstract 51 FHE Fuel Handling Equipment System 52 RBC Reactor Building Cranes 53 PSS Process Sampling System 54 EHVS 13.8 kV and Switchyard System 55 EMVS Medium Voltage AC Electrical Distribution System 56 ELVS Low Voltage AC Electrical Distribution System 57 EDSS Highly Reliable DC Power System 58 EDNS Normal DC Power System 59 BPSS Backup Power Supply 60 PLS Plant Lighting System 61 MCS Module Control System 62 PCS Plant Control System 63 MPS Module Protection System 64 PPS Plant Protection System 65 NMS Neutron Monitoring System 66 SDIS Safety Display and Indication 67 RMS Fixed Area Radiation Monitoring System 68 COMS Communication System 69 SMS Seismic Monitoring System 70 HFT Hot Functional Testing 71 MAEB Module Assembly Equipment Bolting 72 SG Steam Generator Flow-Induced Vibration 73 N/A Security Access Control 74 N/A Security Detection and Alarm 75 N/A Initial Fuel Loading Precritical 76 N/A Initial Fuel Load 77 N/A Reactor Coolant System Flow Measurement 78 N/A NuScale Power Module Temperatures 79 N/A Primary and Secondary System Chemistry 80 N/A Control Rod Drive System-Manual Operation, Rod Speed, and Rod Position Indication 81 N/A Control Rod Assembly Drop Time 82 N/A Pressurizer Spray Bypass Flow 83 N/A Initial Criticality 84 N/A Post-Critical Reactivity Computer Checkout 85 N/A Low Power Test Sequence 86 N/A Determination of Zero-Power Physics Testing Range 87 N/A All Rods Out Boron Endpoint Determination 88 N/A Isothermal Temperature Coefficient Measurement 89 N/A Bank Worth Measurement 90 N/A Power-Ascension 91 N/A Core Power Distribution Map 92 N/A Nuclear Monitoring System Power Range Flux Calibration 93 N/A Reactor Coolant System Temperature Instrument Calibration 94 N/A Reactor Coolant System Flow Calibration 95 N/A Radiation Shield Survey 96 N/A Reactor Building Ventilation System Capability 97 N/A Thermal Expansion 98 N/A Control Rod Assembly Misalignment 99 N/A Steam Generator Level Control 100 N/A Ramp Change in Load Demand Tier 2 14.2-202 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-109: List of Test Abstracts (Continued)

Test Number System Abbreviation Test Abstract 101 N/A Step Change in Load Demand 102 N/A Loss of Feedwater Heater 103 N/A 100 Percent Load Rejection 104 N/A Reactor Trip from 100 Percent Power 105 N/A Island Mode Test for NuScale Power Module #1 106 N/A Island Mode Test for Multiple NuScale Power Modules 107 N/A Remote Shutdown Workstation 108 N/A NuScale Power Module Vibration Test Tier 2 14.2-203 Revision 1

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-110: ITP Testing of New Design Features New System or Component Design Design Feature Tested in the Initial FSAR Section 14.2 Test Program Test Number Containment isolation valves

valve leak rate test #431

valve response to manual ESF action at #636 hot functional test pressure and temperature

valve response time test at hot #637 functional test pressure and temperature

valve response to manual reactor trip #104 at 100% power ECCS valve design

valve response to manual ESF action at #636 hot functional test pressure and temperature

test of valve inadvertent actuation block at design pressure DHRS valve design

valve response to manual ESF action at #636 hot functional test pressure and temperature

valve response to manual reactor trip #104 at 100% power DHRS heat exchanger design

heat exchanger response to manual #104 reactor trip at 100% power Containment flooding and drain system

automatic fill of containment #42

automatic drain of containment Containment evacuation system

establish and maintain containment #41 vacuum

provide RCS leakage detection CNTS level sensors

provides containment level input for #42 CFDS automatic fill and drain of containment RCS flow sensors

provides RCS flow indication during #77 HFT and power ascension testing #94 Pressurizer level sensors

Provides input for pressurizer level #38-1 control Island mode operation

NuScale Power Modules can operate #105 and #106 independently from offsite transmission grid.

Tier 2 14.2-204 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria 14.3 Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria 14.3.1 Introduction This section provides guidance regarding the development of certified design material (CDM) in Tier 1, including Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) required under Title 10 of the Code of Federal Regulations (10 CFR) 52.47(b)(1). The scope of ITAAC is sufficient to provide reasonable assurance that, if the ITAAC are successfully completed, the facility has been constructed and can be operated in accordance with the Atomic Energy Act, relevant Nuclear Regulatory Commission (NRC) regulations, and the combined license (COL). The successful completion of ITAAC constitutes the basis for the NRC determination to allow operation of a facility certified under 10 CFR 52.

Tier 1 information is the portion of the design-related information contained in the Final Safety Analysis Report that is approved and certified by the design certification rule. There are two material categories in Tier 1, the CDM and ITAAC. The CDM is in the form of design descriptions, design commitments, tables, and figures, and is binding for the lifetime of a facility. The ITAAC is used to verify the as-built design features. The ITAAC material expires at initial fuel loading.

The Tier 1 design description consists of the system description and design commitments, both of which contain top-level design features. A design feature is either a physical attribute or a performance characteristic of structures, systems, and components (SSC).

The design features in the system description are not verified by ITAAC. Only the design features in the design commitments are verified by ITAAC.

The sections below describe the criteria and methods by which specific technical entries for Tier 1 were selected. The contents of Tier 1 may not directly correspond to these guidelines in all cases because special considerations may warrant a different approach. In this regard, a case-by-case determination is made consistent with the principles inherent in 10 CFR 52 as well as NRC guidance regarding the content of design descriptions and ITAAC.

COL Item 14.3-1: A COL applicant that references the NuScale Power Plant design certification will provide the site-specific selection methodology and inspections, tests, analyses, and acceptance criteria for emergency planning.

COL Item 14.3-2: A COL applicant that references the NuScale Power Plant design certification will provide the site-specific selection methodology and inspections, tests, analyses, and acceptance criteria for structures, systems, and components within their scope.

14.3.2 Tier 1 Design Description and Inspections, Tests, Analyses, and Acceptance Criteria First Principles General criteria that provide clarity on the scope and level of detail of design descriptions and ITAAC are discussed below. These criteria are consolidated and grouped into two sets of first principles: 1) Tier 1 design description scope first principles and 2) ITAAC scope first principles.

Tier 2 14.3-1 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria The development of the scope of Tier 1 is determined based upon policies set forth by the NRC. A "first principles" approach is considered such that the design descriptions and ITAAC in Design Certification Applications are "necessary and sufficient." Thus, in order to determine the appropriate scope of ITAAC, it is important to apply both the first principles for determining the top-level design features that are included in Tier 1 design descriptions, and the first principles for determining whether a Tier 1 design description needs an ITAAC. Consistent with these first principles, the selection of the top-level design features for Tier 1 is based on the safety significance of SSC, their importance in various safety analyses, and their functions for defense-in-depth considerations.

The first principles for determining the scope of design descriptions and ITAAC are described in Section 14.3.2.1 and Section 14.3.2.2.

14.3.2.1 Tier 1 Design Description Scope First Principles

Design description content is limited to the following:

top-level design features of safety-related SSC top-level design features of safety-related or nonsafety-related SSC that protect safety-related components top-level design features of security system physical SSC top-level design features of risk-significant, nonsafety-related SSC determined by results of a probabilistic risk assessment (PRA)

Refer to Section 14.3.2.1.1 for further discussion of this principle.

Design descriptions are derived solely from Tier 2 design information.

The amount of detail in design descriptions is proportional to the safety significance of the system (i.e., a graded approach). Refer to Section 14.3.2.1.2 for further discussion of the graded approach.

Not all safety-related design features are included in a design description. Refer to Section 14.3.2.1.3 for further discussion of this principle.

Not all design features contained in the accident analyses must be included in design descriptions. Refer to Section 14.3.2.1.4 for further discussion of this principle.

Operational programs and post-fuel load testing are not contained in design descriptions. Refer to Section 14.3.2.1.5 for further discussion of this principle.

Design descriptions do not need to include every component for that system, but instead only includes those SSC that are required to perform the safety-related and risk-significant system functions. Refer to Section 14.3.2.1.6 for further discussion of this principle.

Some risk-significant design features identified by the PRA do not need to be specifically addressed in the design description because they are indirectly addressed by design features that are addressed by other design commitments.

Refer to Section 14.3.2.1.7 for further discussion of this principle.

Tier 2 14.3-2 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria

To the extent that an SSC is already the subject of a design commitment by reason of design basis accident (DBA) mitigation function, a design commitment does not need to address the function of the SSC to mitigate severe accidents. Other design features that are not specifically installed for severe accident mitigation, but are used for severe accident mitigation do not need to be addressed in Tier 1. Refer to Section 14.3.2.1.8 for further discussion of this principle.

Design descriptions only include fixed design features that are installed prior to fuel loading and are expected to be in place for the lifetime of the plant. Refer to Section 14.3.2.1.9 for further discussion of this principle.

No new design information can be contained in Tier 1 that is not already in Tier 2.

Tier 1 information is not relied upon for the NRC safety determination provided in a Safety Evaluation Report. The NRC safety determination is based solely on the Tier 2 design information. If a system or component function or design feature is not discussed in Tier 2, hence is not part of the NRC safety determination, it does not belong in Tier 1.

Design descriptions do not contain a level of detail (e.g., minor dimensional details) that would restrict a licensee from making changes that do not affect a safety-related or risk-significant system function.

Systems with no safety significance are not included in design descriptions.

Design descriptions do not contain information that the NRC may designate as "Tier 2*". The NRC guidance in NUREG-0800, Section 14.3, states that Tier 2*

information is generally not appropriate for treatment in Tier 1 because it is subject to change.

Design descriptions do not contain processes that are used for designing and constructing a plant because the safety-related function of an SSC is dependent upon its final as-built condition and not the processes used to achieve that condition.

Design descriptions do not contain discussions of single failures. Rather, the design description contains related top-level design features, such as physical separation and electrical isolation of Class 1E circuits.

14.3.2.1.1 Tier 1 Design Descriptions Are Limited to the Top-Level Design Features The following describes the top-level design features for the NuScale Power Plant.

A design feature is either a physical attribute or a performance characteristic of an SSC. The top-level design features contained in Tier 1 design descriptions are:

reactor coolant pressure boundary

containment pressure boundary

Seismic Category I Reactor Building (RXB) and Control Building (CRB)

Radwaste Category RW-IIa Radioactive Waste Building (RWB)

control room envelope (CRE)

safety-related equipment qualification Tier 2 14.3-3 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria

safety-related component performance

SSC providing protection of safety-related components

safety-related protection system (reactor trip and engineered safety features actuation systems (ESFAS))

components providing radiation protection for personnel and safety-related equipment

new and spent fuel storage

security system physical components Examples of structures included in Tier 1 design descriptions are the Seismic Category I RXB and CRB, fire barriers, flood barriers, radiation shields, and fuel storage racks.

Examples of components included in Tier 1 design descriptions are valves, instruments, and piping systems.

Examples of physical attributes included in Tier 1 design descriptions are safety-related equipment qualification, location of fire barriers, and thickness of radiation shields.

Examples of performance characteristics included in Tier 1 design descriptions are building seismic performance, safety-related piping conformance to American Society of Mechanical Engineers (ASME) Code Section III requirements, valve stroke time, and safety-related components' automatic response to the module protection system (MPS).

14.3.2.1.2 Graded Approach The extent to which a particular SSC is described in Tier 1 depends upon the safety significance of the SSC. A graded approach is used to determine the type of information and the level of detail in Tier 1 commensurate with the safety significance of the SSC for the design.

The graded approach reflects the wide variation in safety significance from system to system. It is unnecessary and would be inappropriate to provide the same level of detail for every system in Tier 1.

Top-level design information in Tier 1 is extracted from the more detailed design information presented in Tier 2. Limiting the Tier 1 contents to top-level information reflects the graded approach consistent with NRC guidance in NUREG-0800 and in Regulatory Guide (RG) 1.206.

Severe accident design features are described in the design description, and the ITAAC verify that they exist. In general, the capabilities of the design features need not be included in the ITAAC. For example, a design commitment may discuss that a severe accident containment flooding system exists, while the acceptance criteria Tier 2 14.3-4 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria would discuss that the severe accident containment flooding system exists, but would not specify the capabilities of associated pumps.

14.3.2.1.3 Not all Safety-Related Design Features are Included in a Design Description Not all safety-related design features need to be explicitly addressed in design descriptions. Examples of safety-related component design features that generally do not warrant discussion in a design description include:

instrument lines

fill lines

drains

ASME Code Section III valves that have only a passive function

piping pressure relief valves associated with thermal expansion and anticipated valve leakage

interlocks aimed specifically at equipment protection for safety-related components

local controls for safety-related components

rebar and concrete properties for Seismic Category I structures 14.3.2.1.4 Top-Level Design Features Not all design features are included in the design descriptions. Only the top-level design features are contained in the appropriate design description and verified by ITAAC. Table 14.3-1 and Table 14.3-2 present a matrix which correlates the top-level design features contained in design commitments with their treatment in Tier

1. Table 14.3-1 and Table 14.3-2 also contains the top-level design features that were developed based upon results of the following plant safety analyses:

transient and accident analyses

internal and external hazards analyses

radiological analyses

risk-significant design features as determined by the results of a PRA

design features necessary or important to severe accident mitigation

fire protection By capturing the top level design features that are based upon results of plant safety analyses, the integrity of the fundamental analyses associated with the design as presented in Tier 2 are preserved in the certified design as presented in Tier 1.

Tier 2 14.3-5 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria 14.3.2.1.5 Design Descriptions do not Include Operational Programs and Post-Fuel Load Testing Those aspects of the design that pertain to programs rather than the as-built plant (e.g., Appendix B to 10 CFR Part 50 requires a quality assurance program, and 10 CFR 50.65 requires a maintenance rule program) are not included in Tier 1.

The key aspects of the design are described in Tier 1. Those aspects of the design that cannot be verified until after fuel loading are not included in ITAAC. This is because 10 CFR 52 requires the ITAAC to be satisfied prior to fuel loading. For these, the Initial Test Program verifies various aspects of the design after fuel load, but prior to operation. Examples are the post-fuel load startup and power ascension test program verification of fuel, control rod, and core characteristics, as well as system and integrated plant operating characteristics. The treatment of these issues is similar to their treatment at facilities licensed under 10 CFR 50, in that verification of the satisfactory completion of these requirements are a condition of the license.

14.3.2.1.6 Design Commitments only include Components Required to Perform System Functions in the System Description Not every design element specified in the certified design rule has a corresponding Tier 1 verification requirement. For example the safety classification of SSC are identified in the design descriptions, but are not verified by ITAAC because there is no specific test for this characteristic. Further, some ITAAC verify system function and do not address individual system components that together yield the required system functional performance.

14.3.2.1.7 Risk-Significant Design Features as Determined by the Results of a Probabilistic Risk Assessment Some risk-significant design features identified by the PRA do not need to be specifically addressed in the design description because they are indirectly addressed by design features that are addressed by other design commitments.

For example, some PRA studies are dependent upon an assessment of the ability of certain SSC to function during seismic events that are more severe than the design basis safe shutdown earthquake (SSE). If equipment is designed and qualified for the seismic design basis, the design process is such that the added capability assumed in the PRA will inherently be present.

The risk-significant design features that are included in the design descriptions and have associated ITAAC are listed in Table 14.3-1 and Table 14.3-2.

14.3.2.1.8 Design Features Necessary or Important to Severe Accident Mitigation There are some SSC that mitigate DBAs as well as provide an important success path for severe accident mitigation. The severe accident analysis design features that are included in the design descriptions and have associated ITAAC are listed in Table 14.3-1 and Table 14.3-2.

Tier 2 14.3-6 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria 14.3.2.1.9 Design Descriptions Only Include Fixed Design Features Installed Prior to Fuel Loading and Expected to be in Place for the Lifetime of the Plant Those aspects of the design that pertain to portable items or consumables rather than fixed design features are not included in Tier 1. Because hardware such as fuel cannot be installed in the reactor until after completion of the ITAAC and because the fuel will be periodically replaced, fuel is not an appropriate topic for ITAAC.

14.3.2.2 Inspections, Tests, Analyses, and Acceptance Criteria Scope First Principles The following criteria are considered when determining which information warrants inclusion in the ITAAC entries:

The design commitment is extracted directly from the design descriptions and differences in text are minimized, unless intentional.

The NRC safety determination is based solely on the Tier 2 design information.

ITAAC are not relied upon for the NRC safety determination provided in a Safety Evaluation Report.

The ITAAC are an important part of the NRC construction verification program, but do not verify every design and construction feature included in the certified design.

The ITAAC are not meant to be a one-for-one check of detailed design and construction features that are verified by the normal construction quality programs.

An inspection, test, or analysis, or a combination thereof, may verify one or more provisions in the design commitment, as defined by the ITAAC.

14.3.3 Organization of Tier 1 The design descriptions, interface requirements, and site parameters are derived from Tier 2 information. Tier 1 information includes

preamble material which includes a table of contents, a list of tables, and a list of figures.

an introduction section (described in Section 14.3.4).

unit-specific design descriptions and ITAAC (described in Section 14.3.5). This section includes systems that are fully within the scope of the NuScale Power Plant design certification.

The in-scope portion of those systems that are only partially within the scope of the NuScale Power Plant design certification.

shared or common SSC and non-SSC design descriptions and ITAAC (described in Section 14.3.6).

interface material (described in Section 14.3.7).

site parameters (described in Section 14.3.8).

Tier 2 14.3-7 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria 14.3.3.1 Design Descriptions (Certified Design Material)

The design descriptions serve as requirements for the lifetime of a plant to assure that the plant does not deviate from the certified design. The design descriptions use a system-based structure that is different than the structure of Tier 2. Consequently, developing the design description entries for a system is based on multiple Tier 2 chapters having technical information related to that system.

The design description consists of a system description and design commitments.

System description tables and figures are used where appropriate.

The top-level design features in Tier 1 are extracted from the more detailed design information in Tier 2 using the first principles described in Section 14.3.2.1.

System Description and Design Commitments The purpose of the system description is to provide a concise description of the safety-related and risk-significant system functions, safety classification, and general location.

The system description only describes those portions of the system that perform safety-related and risk-significant functions.

The level of detail in system descriptions uses a graded approach commensurate with the safety and risk significance of a system.

Design commitments are provided in numbered paragraphs that are used to develop the design commitment column in the ITAAC table as discussed in Section 14.3.2.2.

Design commitments cover design features, such as seismic and ASME Code classifications, Class 1E power sources and divisions, equipment to be qualified for harsh environments (and other than harsh for certain instrumentation and controls (I&C) equipment).

System Description Tables A table may be used in cases where portions of the system description can be more concisely presented in tabular form. System description tables are generally only referenced in the ITAAC acceptance criteria. System description tables are used to identify design features such as ASME Code class, valve active functions, Class 1E classification of electrical equipment, or required response time of equipment.

System Description Figures A figure may be included in Tier 1 if it is necessary to describe something that cannot be adequately described in the system description and tables. Figures are provided to convey information in support of system descriptions, in cases where information can be more concisely presented in a figure. Figures are intended to depict a simplified schematic arrangement of the significant SSC.

Tier 2 14.3-8 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria 14.3.3.2 Inspections, Tests, Analyses, and Acceptance Criteria Tables A table of ITAAC entries is provided for each system that has design commitments in the design description. A three-column format for the ITAAC table is used. All three columns of the ITAAC table must be read and interpreted together.

The first column of the ITAAC table identifies the design commitment to be verified.

This column contains the specific text of the design commitment, which is extracted from the design commitments contained in the design description.

The second column of the ITAAC table identifies the proposed method by which the licensee will verify the design commitment described in column 1. The methods used are inspections, tests, analyses, or a combination of the three.

Inspections are used when verification can be done by visual observation, physical examination, or reviews of records based on visual observation or physical examination that compare a) the SSC condition to one or more design commitments or b) the program implementation elements to one or more program commitments, as applicable. Examples include walkdowns, configuration checks, measurements of dimensions, or nondestructive examinations.

Tests mean actuation or operation, or establishment, of specified conditions to evaluate the performance or integrity of as-built SSC, unless explicitly stated otherwise, to determine whether an ITAAC acceptance criterion is met.

In addition to testing equipment at its final location, alternative testing methods may be used including factory testing, test facility testing, and laboratory testing.

Testing can also include type testing such as might be performed to demonstrate qualification to meet environmental requirements. Type test means a test on one or more sample components of the same type and manufacturer to qualify other components of the same type and manufacturer. A type test is not necessarily a test of an as-built SSC.

Analyses are used when verification can be done by calculation, mathematical computation, or engineering or technical evaluations.

The third column of the ITAAC table identifies the specific acceptance criteria for the inspections, tests, or analyses described in column 2 that, if met, demonstrate that the licensee has met the design commitments in column 1. Acceptance criteria are objective and clear to avoid confusion over whether or not acceptance criteria have been satisfied.

Using the criteria listed above, ITAAC table entries were developed for each selected system. This was achieved by evaluating the design features defined in the design descriptions and preparing an ITAAC table entry for each design description entry that satisfied the above selection criteria.

Having established the design features for which ITAAC are appropriate, the ITAAC table was completed by selecting the method to be used for verification (either an inspection, a test, or an analysis, or a combination of these) and the acceptance criteria Tier 2 14.3-9 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria against which the as-built design features are measured. The proposed verification activity is identified in the second column of the ITAAC table.

Where ITAAC is verified by a preoperational test, the test will be established in accordance with the Initial Test Program described in Section 14.2 and RG 1.68.

Conversion or extrapolation of test results from the test conditions to design condition may be necessary to satisfy specific ITAAC.

Selection of acceptance criteria is dependent upon the specific design characteristic being verified by the ITAAC table entry. In most cases the appropriate acceptance criteria are self- evident and are based upon the design descriptions. For many of the ITAAC, the acceptance criterion is a statement that the as-built facility has the design feature identified in the design description. A guiding principle for acceptance criteria preparation is the recognition that the criteria should be objective and unambiguous.

In some cases, the ITAAC contain numerical values from Tier 2 that are not specifically identified in the design description or the design commitment column of the ITAAC table. This is acceptable because the design description defines the important design feature that merits Tier 1 treatment. The numerical value in the acceptance criterion is a measurement standard for determining if the as-built facility is in compliance with the design commitment.

The use of objective and unambiguous terms for the acceptance criteria minimizes opportunities for multiple, subjective (and potentially conflicting) interpretations as to whether an acceptance criterion has, or has not, been met. In some cases, the acceptance criteria may be more general because the detailed supporting information in Tier 2 does not lend itself to concise verification. Numerical values for SSC are specified as ITAAC acceptance criteria when values consistent with the design commitments are possible, or when failure to meet the stated acceptance criterion would clearly indicate a failure to properly implement the design or meet the safety analysis.

Where appropriate, the detailed design information provided in Tier 2 includes supporting information for various inspections, tests, and analyses that is used to satisfy the acceptance criteria. This information describes an acceptable means of satisfying an ITAAC.

The details in Tier 2 are not referenced in Tier 1 and are not part of the CDM.

For numerical values in the acceptance criteria, ranges or tolerances are generally included. This is necessary and acceptable because:

Specification of a single-value acceptance criterion is impractical because minute deviations would represent noncompliance.

Tolerances recognize that legitimate site variations can occur in complex construction projects.

Minor variations in plant parameters within the tolerance bounds have no effect on plant safety.

Tier 2 14.3-10 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria 14.3.3.3 Systems Within the Scope of Tier 1 The results of the ITAAC screenings of SSC that are either fully or partially within the scope of the NuScale Power Plant design certification are provided in Table 14.3-1 and Table 14.3-2. These tables identify those SSC that are addressed in Tier 1.

Tier 1 does not include systems that have been determined to not require design descriptions or ITAAC.

14.3.4 Tier 1 Chapter 1, Introduction Tier 1 Chapter 1 contains the definitions and general provisions used in design descriptions and ITAAC. The intent of these entries is to avoid ambiguities and misinterpretations by providing front-end guidance to users of Tier 1.

Definitions are included for terms used in Tier 1 that could be subject to various interpretations. The intent is to be consistent with Tier 2 information and to reflect NRC guidance regarding various terms. Should questions on terminology arise, the definitions would aid in understanding the intent of the information in Tier 1.

General provisions are included for treatment of individual items, implementation of ITAAC (including ITAAC format), discussion of matters related to operations, and interpretation of figures. The rated reactor core thermal power is not specified because the maximum power level with any special conditions will be specified in the operating license.

Tier 2 Table 1.1-1 is used to interpret Tier 1. The information in Table 1.1-1 will not be duplicated in Tier 1 Chapter 1 in order to prevent the treatment of acronyms and abbreviations as Tier 1 CDM.

The figure legend contained in Tier 2 Chapter 1 is used to interpret Tier 1 system description figures. The information in Figures 1.7-1 through 1.7-3 will not be duplicated in Tier 1 Chapter 1 in order to prevent the treatment of figure legends as Tier 1 CDM.

14.3.5 Tier 1 Chapter 2, Unit-Specific Structures, Systems, and Components Design Descriptions and Inspections, Tests, Analyses, and Acceptance Criteria Tier 1 Chapter 2 contains design descriptions and associated ITAAC for unit-specific systems that support a single NuScale Power Module (NPM). The unit-specific system design is identical between units. If a unit-specific system meets the first principles for entry into Tier 1 as described in Section 14.3.2.1, then its design description and ITAAC are entered into Tier 1. Tier 1 Chapter 2 includes an entry for each unit-specific system that is either fully or partially within the scope of the NuScale Power Plant design certification as identified in Table 14.3-1.

The design descriptions of a given unit-specific system are the same for all units.

However, unlike single-unit facility designs, each ITAAC for a given unit-specific system must be completed for each unit.

Tier 2 14.3-11 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria The design description for a unit-specific system will only be recorded once in Tier 1, but the ITAAC for that system must be completed for each unit.

14.3.6 Tier 1 Chapter 3, Shared Structures, Systems, and Components and Non-Structures, Systems, and Components Design Descriptions and Inspections, Tests, Analyses, and Acceptance Criteria Tier 1 Chapter 3 contains design descriptions and associated ITAAC for systems that support multiple NPMs (shared or common systems). If a shared or common system meets the first principles for entry into Tier 1 as described in Section 14.3.2.1, then its design description and ITAAC are entered into Tier 1. Tier 1 Chapter 3 includes an entry for the shared or common systems that are either fully or partially within the scope of the NuScale Power Plant design certification as identified in Table 14.3-2. Additionally, Tier 1 Chapter 3 addresses non-SSC design and construction activities that are applicable to more than one system or NPM such as human factors engineering.

Shared or common systems that must be completed to support the operation of the first NPM have their ITAAC completed once. If shared or common systems require a portion of the system to be completed to support the operation of the first NPM, then the applicable ITAAC will be in Tier 1 Chapter 2 and must be completed for each associated NPM.

Entries in this chapter of Tier 1 have the same structure as the unit-specific material discussed in Section 14.3.5; that is, design description text, tables, figures, and a table of ITAAC entries.

14.3.7 Tier 1 Chapter 4, Interface Requirements Tier 1 Chapter 4 provides the interface requirements. Interface requirements are design features that are met by the site-specific portions of a facility that are not within the scope of the certified design. The interface requirements define the design features that ensure the site-specific portion of the design is in conformance with the certified design. The site-specific portions of the design are those portions of the design that are dependent on characteristics of the site.

Tier 1 Chapter 4 also identifies the scope of the design to be certified by specifying the systems that are completely or partially out of scope of the certified design. Thus, interface requirements are defined for: (a) systems that are entirely outside the scope of the certified design, and (b) the out-of-scope portions of those systems that are only partially within the scope of the certified design.

The NuScale Power Plant relies upon passive safety features physically located within NuScale Power Plant buildings and structures. No interfaces need to be identified between or among these portions of the facility. Tier 1 Chapter 4 does not include ITAAC or a requirement for COL developed ITAAC for interface requirements.

14.3.8 Tier 1 Chapter 5, Site Parameters Tier 1 Chapter 5 provides bounding values for site parameters that a COL applicant referencing the NuScale Power Plant design certification will use in the design of a specific site. Compliance with these site parameters is verified during the COL application process, Tier 2 14.3-12 Revision 1

Certified Design Material and Inspections, Tests, Analyses, and NuScale Final Safety Analysis Report Acceptance Criteria so no ITAAC are necessary for site parameters. Chapter 2 provides a discussion of the envelope of site design parameters used for the NuScale Power Plant design. The corresponding Tier 1 Chapter 5 is based on Table 2.0-1. Tier 1 Chapter 5 is limited to a tabular entry; no supporting text material is required.

Tier 2 14.3-13 Revision 1

Tier 2 NuScale Final Safety Analysis Report Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.01 NPM As required by ASME Code Section III NCA-1210, each ASME Code X Class 1, 2 and 3 component (including piping systems) of a nuclear power plant requires a Design Report in accordance with NCA-3550. NCA-3551.1 requires that the drawings used for construction be in agreement with the Design Report before it is certified and be identified and described in the Design Report. It is the responsibility of the N Certificate Holder to furnish a Design Report for each component and support, except as provided in NCA-3551.2 and NCA-3551.3. NCA-3551.1 also requires that the Design Report be certified by a registered professional engineer when it is for Class 1 components and supports, Class CS core support structures, Class MC vessels and supports, Class 2 vessels designed to NC-3200 (NC-3131.1), or Class 2 or Class 3 components designed 14.3-14 to Service Loadings greater than Design Loadings. A Class 2 Design Report shall be prepared for Class 1 piping NPS 1 or smaller that is Certified Design Material and Inspections, Tests, Analyses, and designed in accordance with the rules of Subsection NC. NCA-3554 requires that any modification of any document used for construction, from the corresponding document used for design analysis, shall be reconciled with the Design Report.

An ITAAC inspection is performed of the NuScale Power Module ASME Code Class 1, 2 and 3 as-built piping system Design Report to verify that the requirements of ASME Code Section III are met.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.02 NPM The ASME Code Section III requires that documentary evidence be X X available at the construction or installation site before use or installation to ensure that ASME Code Class 1, 2 and 3 components conform to the requirements of the Code. As defined in NCA-9000, a component can be a vessel, pump, pressure relief valve, line valve, storage tank, piping system, or core support structure that is designed, constructed, and stamped in accordance with the rules of Section III. The NuScale Power Module ASME Code Class 1 and 2 components require a Data Report as specified by NCA-1210. The Data Report is prepared by the certificate holder or owner and signed by the certificate holder or owner and the inspector as specified by NCA-8410. The type of individual Data Report forms necessary to record the required code data is specified in Table NCA-8100-1.

An ITAAC inspection is performed of the Data Reports for NuScale 14.3-15 Power Module ASME Code Class 1 and 2 as-built components listed Certified Design Material and Inspections, Tests, Analyses, and in Tier 1 Table 2.1-2 and interconnecting piping to (1) ensure that the appropriate Data Reports have been provided as specified in Table NCA-8100-1, (2) ensure that the certificate holder or owner and the authorized nuclear inspector have signed the Data Reports, and (3) verify that the requirements of ASME Code met.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.03 NPM The ASME Code Section III requires that documentary evidence be X available at the construction or installation site before use or installation to ensure that ASME Code CS components conform to the requirements of the Code. The ASME Code Class CS components require a Data Report as specified by NCA-1210. The Data Report is prepared by the certificate holder or owner and signed by the certificate holder or owner and the Inspector as specified by NCA-8410. The type of individual Data Report Forms necessary to record the required Code Data is identified in Table NCA-8100-1.

An ITAAC inspection is performed of the Data Reports for the ASME Code Class CS as-built components listed in Tier 1 Table 2.1-2 to (1) ensure that the appropriate Data Reports have been provided as specified in Table NCA-8100-1, (2) ensure that the certificate holder or owner and the inspector have signed the Data Reports, and (3) 14.3-16 verify that the requirements of ASME Code Section III are met.

Certified Design Material and Inspections, Tests, Analyses, and 02.01.04 NPM Section 3.6, Protection against Dynamic Effects Associated with X X Postulated Rupture of Piping, provides the design bases and criteria for the analysis required to demonstrate that safety-related SSC are not impacted by the adverse effects of a high-and moderate-energy pipe failure within the plant. Table 3.6-2:

Postulated Break Locations, lists the high-and moderate-energy pipe break locations.

An ITAAC inspection is performed to verify that the as-built protective features credited in the reconciled Pipe Break Hazards Analysis Report such as pipe whip restraints, pipe whip or jet impingement barriers, jet impingement shields, or guard pipe have Acceptance Criteria been installed in accordance with design drawings of sufficient detail to show the existence and location of the protective hardware. The as-built inspection is intended to verify that changes to postulated pipe failure locations and protective features or protected equipment made during construction do not adversely affect the safety-related functions of the protected equipment.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.05 NPM Section 3.6.3, Leak-Before-Break Evaluation Procedures, describes X the application of the mechanistic pipe break criteria, commonly referred to as leak-before-break (LBB), to the evaluation of pipe ruptures. The LBB analysis eliminates the need to consider the dynamic effects of postulated pipe breaks for high-energy piping that qualify for LBB.

An analysis, which includes material properties of piping and welds, stress analyses, leakage detection capability, and degradation mechanisms, confirms that the as-designed LBB analysis is bounding for the ASME Code Class 2 as-built piping listed in Tier 1 Table 2.1-1 and interconnected equipment nozzles.

A summary of the results of the plant specific LBB analysis, including material properties of piping and welds, stress analyses, leakage detection capability, and degradation mechanisms is provided in the as-built LBB analysis report.

14.3-17 02.01.06 NPM Section 5.3.1.5, Fracture Toughness, discusses the fracture X Certified Design Material and Inspections, Tests, Analyses, and toughness properties of the reactor pressure vessel (RPV) beltline material and the Material Surveillance Program. A Charpy V-Notch test of the RPV beltline material specimen is performed by the vendor to ensure that the initial RPV beltline Charpy upper-shelf energy is no less than 75 ft-lb.

02.01.07 NPM Section 6.2.6, Containment Leakage Testing, provides a discussion X of the leakage testing requirements of the containment vessel (CNV), which serves as an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment. As discussed in Section 6.2.6, the NuScale CNV is exempted from the integrated leak rate testing specified in the General Design Criterion (GDC) 52.

Acceptance Criteria In accordance with Table 14.2-43, a preoperational test demonstrates that the leakage rate for local leak rate tests (Type B and Type C) for pressure containing or leakage-limiting boundaries and containment isolation valves (CIVs) meet the leakage Revision 1 acceptance criterion of 10 CFR Part 50, Appendix J.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.08 NPM Section 6.2.4.3, Design Evaluation, provides a discussion of how the X containment system (CNTS) containment isolation valves close within the required closure time after receipt of a containment isolation signal to meet containment isolation requirements following a radiological release in the CNV.

In accordance with Table 14.2-63, a preoperational test demonstrates that each automatic CIV listed in Tier 1 Table 2.1-3 travels from the full open to full closed position in less than or equal to the time listed in Table 6.2-5 after receipt of a containment isolation signal.

02.01.09 NPM Section 6.2.4.2.2, Component Description, provides a discussion of X the isolation valves outside containment that are located as close to the containment as practical in accordance with the requirements of 10 CFR Part 50, Appendix A, GDC 55, 56 and 57.

14.3-18 An ITAAC inspection is performed to verify the length of piping Certified Design Material and Inspections, Tests, Analyses, and between each containment penetration and its associated outboard CIVs is less than or equal to the length identified in Tier 1 Table 2.1-1.

02.01.10 NPM Section 8.1.5.3 General Design Criteria, NRC Regulations, RGs, and X Branch Technical Positions, NUREG Reports, SECY Papers, and NRC Bulletins, discusses that the NPM CNTS containment electrical penetration assemblies are sized to power their design loads as demonstrated by satisfying the guidance of RG 1.63.

An analysis determines the required design electrical rating needed to power the design loads of each NPM CNTS containment electrical penetration assembly listed in Tier 1 Table 2.1-3.

Acceptance Criteria An ITAAC inspection is performed to verify that the electrical rating of each NPM CNTS containment electrical penetration assembly listed in Tier 1 Table 2.1-3 is greater than or equal to the required design electrical rating. This ITAAC inspection may be performed any time after manufacture of the CNTS containment electrical Revision 1 penetration assemblies.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.11 NPM Sections 7.1.2, Independence, discusses the independence of the X MPS Class 1E instrumentation and control current-carrying circuits per the guidance of RG 1.75, which endorses Institute of Electrical and Electronics Engineers (IEEE) Std. 384-1992. Physical separation is provided to maintain the independence of Class 1E I&C current-carrying circuits so that the safety functions required during and following any design basis event can be accomplished. Minimum separation distance (as defined in IEEE Std. 384-1992), or barriers or any combination thereof may achieve physical separation as specified in IEEE Std. 384-1992.

Separate ITAAC inspections are performed to verify the independence provided by physical separation and the independence provided by electrical isolation. This ITAAC verifies the independence of Class 1E current-carrying circuits by physical separation. The scope of this commitment includes the cables from 14.3-19 the NPM disconnect box to the instrument. An ITAAC inspection is Certified Design Material and Inspections, Tests, Analyses, and performed of physical separation of the MPS Class 1E current-carrying circuits. The physical separation ITAAC inspection results verify that the following physical separation criteria are met:

i. Physical separation between redundant divisions of the MPS Class 1E I&C current-carrying circuits is provided by a minimum separation distance, or by barriers (where the minimum separation distances cannot be maintained), or by a combination of separation distance and barriers; and such physical separation satisfies the criteria of RG 1.75. The configuration of each as-built barrier agrees with its associated as-built drawing.

ii. Physical separation between the MPS Class 1E I&C current-Acceptance Criteria carrying circuits and non-Class 1E I&C current-carrying circuits is provided by a minimum separation distance, or by barriers (where the minimum separation distances cannot be maintained), or by a combination of separation distance and barriers; and such physical separation satisfies the criteria of RG 1.75. The configuration of Revision 1 each as-built barrier agrees with its associated as-built drawing.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.12 NPM Section 5.3.1.6, Material Surveillance, discusses the use of specimen X capsules installed in specimen guide baskets.

An ITAAC inspection is performed to verify that the correct number of guide baskets are attached to the outer surface of the core barrel at about the mid height of the core support assembly at approximately 90-degree intervals.

02.01.13 NPM The CNTS remotely operated CNTS containment isolation valves X are tested by remote operation to demonstrate the capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-63, a preoperational test demonstrates that the CNTS remotely operated CNTS containment isolation valves listed in Table 2.1-2 stroke fully open and fully 14.3-20 closed by remote operation under preoperational test conditions.

Certified Design Material and Inspections, Tests, Analyses, and Preoperational test conditions are established that approximate design-basis temperature, differential pressure, and flow conditions to the extent practical, consistent with preoperational test limitations.

02.01.14 NPM The emergency core cooling system (ECCS) safety-related valves X are tested by remote operation to demonstrate the capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-63, a preoperational test demonstrates that the ECCS safety-related valves listed in Table Acceptance Criteria 2.1-2 stroke fully open and fully closed by remote operation under preoperational test conditions.

Preoperational test conditions are established that approximate design-basis temperature, differential pressure, and flow conditions to the extent practical, consistent with preoperational Revision 1 test limitations.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.15 NPM The decay heat removal system (DHRS) safety-related valves are X tested by remote operation to demonstrate the capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-63, a preoperational test demonstrates that the DHRS safety-related valves listed in Table 2.1-2 stroke fully open and fully closed by remote operation under preoperational test conditions.

Preoperational test conditions are established that approximate design basis temperature, differential pressure, and flow conditions to the extent practical, consistent with preoperational test limitations.

02.01.16 NPM The reactor coolant system (RCS) safety-related check valves are X 14.3-21 tested to demonstrate the capability to perform their function to Certified Design Material and Inspections, Tests, Analyses, and transfer open and transfer closed (under forward and reverse flow conditions, respectively) under preoperational temperature, differential pressure, and flow conditions. Check valves are tested in accordance with the requirements of the ASME OM Code, ISTC-5220, Check Valves.

In accordance with Table 14.2-46, a preoperational test demonstrates that the RCS check valves listed in Table 2.1-2 strokes fully open and closed under forward and reverse flow conditions, respectively.

Preoperational test conditions are established that approximate design-basis temperature, differential pressure and flow conditions Acceptance Criteria to the extent practical, consistent with preoperational test limitations.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.17 NPM The RCS safety-related excess flow check valves are tested to X X demonstrate the capability to perform their function to stroke fully closed under excess flow conditions under preoperational temperature, differential pressure, and flow conditions. Check valves are tested in accordance with the requirements of the ASME OM Code, ISTC-5220, Check Valves.

In accordance with Table 14.2-46, a preoperational test demonstrates that the RCS check valves listed in Table 2.1-2 strokes fully closed under forward flow conditions.

Preoperational test conditions are established that approximate design-basis temperature, differential pressure and flow conditions to the extent practicable, consistent with preoperational test limitations.

02.01.18 NPM The CNTS safety-related hydraulic-operated valves are tested to X 14.3-22 demonstrate the capability to perform their function to fail to or Certified Design Material and Inspections, Tests, Analyses, and maintain their safety-related position on loss of motive power under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-63, a preoperational test demonstrates that each CNTS safety-related hydraulic-operated valves listed in Table 2.1-2 repositions to or maintains its safety-related position on loss of motive power (electric power to the valve actuating solenoid(s) is lost, or hydraulic pressure to the valve(s) is lost).

Preoperational test conditions are established that approximate design-basis temperature, differential pressure, and flow Acceptance Criteria conditions to the extent practicable, consistent with preoperational test limitations.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.19 NPM The ECCS safety-related reactor recirculation valves and reactor X X vent valves are tested to demonstrate the capability to perform their function to fail to or maintain their safety-related position on loss of electrical power under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-63, a preoperational test demonstrates that each ECCS safety-related reactor recirculation valve and reactor vent valve listed in Table 2.1-2 fails open on loss of electrical power to its corresponding trip valve.

Preoperational test conditions are established that approximate design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

02.01.20 NPM The DHRS safety-related hydraulic-operated valves are tested to X 14.3-23 demonstrate the capability to perform their function to fail to or Certified Design Material and Inspections, Tests, Analyses, and maintain their safety-related position on loss of motive power under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-63, a preoperational test demonstrates that each DHRS safety-related hydraulic-operated valves listed in Table 2.1-2 fails open loss of motive power (electric power to the valve actuating solenoid(s) is lost, or hydraulic pressure to the valve(s) is lost).

Preoperational test conditions are established that approximate design basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test Acceptance Criteria limitations.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.01.21 NPM The CNTS safety-related check valves are tested to demonstrate the X capability to perform their function to transfer open and transfer closed (under forward and reverse flow conditions, respectively) under preoperational temperature, differential pressure, and flow conditions. Check valves are tested in accordance with the requirements of the ASME OM Code, ISTC-5220, Check Valves.

In accordance with Table 14.2-43, a preoperational test demonstrates that the CNTS check valves listed in Tier 1 Table 2.1-2 strokes fully open and closed under forward and reverse flow conditions, respectively.

Preoperational test conditions are established that approximate design basis temperature, differential pressure and flow conditions to the extent practicable, consistent with preoperational test limitations.

14.3-24 02.01.22 NPM Section 8.3.1.2.2, Circuit Protection and Coordination, discusses X Certified Design Material and Inspections, Tests, Analyses, and instantaneous and thermal overload fault protection to limit the loss of equipment due to postulated fault conditions.

A circuit interrupting device coordination analysis confirms that the as-built containment electrical penetration assemblies listed in Tier 1 Table 2.1-3 can withstand fault currents for the time required to clear the fault from its power source.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.02.01 CVCS As required by ASME Code Section III NCA-1210, each ASME Code X Class 1, 2 and 3 component (including piping systems) of a nuclear power plant requires a Design Report in accordance with NCA-3550. NCA-3551.1 requires that the drawings used for construction be in agreement with the Design Report before it is certified and be identified and described in the Design Report. It is the responsibility of the N certificate holder to furnish a Design Report for each component and support, except as provided in NCA-3551.2 and NCA-3551.3. NCA-3551.1 also requires that the Design Report be certified by a registered professional engineer when it is for Class 1 components and supports, Class CS core support structures, Class MC vessels and supports, Class 2 vessels designed to NC-3200 (NC-3131.1), or Class 2 or Class 3 components designed to service loadings greater than design loadings. A Class 2 Design Report shall be prepared for Class 1 piping NPS 1 or smaller which is 14.3-25 designed in accordance with the rules of Subsection NC. NCA-3554 requires that any modification of any document used for Certified Design Material and Inspections, Tests, Analyses, and construction, from the corresponding document used for design analysis, shall be reconciled with the Design Report.

An ITAAC inspection is performed of the chemical and volume control system (CVCS) ASME Code Class 3 as-built piping system Design Report to verify that the requirements of ASME Code Section III are met.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.02.02 CVCS The ASME Code Section III requires that documentary evidence be X available at the construction or installation site before use or installation to ensure that ASME Code Class 1, 2 and 3 components conform to the requirements of the Code. As defined in NCA-9000, a component can be a vessel, pump, pressure relief valve, line valve, storage tank, piping system, or core support structure that is designed, constructed, and stamped in accordance with the rules of Section III. The chemical and volume control system ASME Code Class 3 components require a Data Report as specified by NCA-1210. The Data Report is prepared by the certificate holder or owner and signed by the certificate holder or owner and the inspector as specified by NCA-8410. The type of individual Data Report forms necessary to record the required code data is specified in Table NCA-8100-1.

An ITAAC inspection is performed of the Data Reports for the 14.3-26 chemical and volume control system ASME Code Class 3 as-built Certified Design Material and Inspections, Tests, Analyses, and components listed in Tier 1 Table 2.2-2 and interconnecting piping that is described in Section 9.3.4 to (1) ensure that the appropriate Data Reports have been provided as specified in Table NCA-8100-1, (2) ensure that the certificate holder or owner and the authorized nuclear inspector have signed the Data Reports, and (3) verify that the requirements of ASME Code Section III are met.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.02.03 CVCS The chemical and volume control ASME Code Class 3 valves are X tested by remote operation to demonstrate the capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

In accordance with the information provided in Table 14.2-38, a preoperational test demonstrates that the chemical and volume control system ASME Code Class 3 valves listed in Tier 1 Table 2.2-2 stroke fully open and fully closed by remote operation under preoperational test conditions.

Preoperational test conditions are established that approximate design basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

14.3-27 02.02.04 CVCS The chemical and volume control system ASME Code Class 3 check X Certified Design Material and Inspections, Tests, Analyses, and valves are tested to demonstrate the capability to perform their function to transfer closed (under reverse flow conditions) under preoperational temperature, differential pressure, and flow conditions. Check valves are tested in accordance with the requirements of the ASME OM Code, ISTC-5220, Check Valves.

In accordance with the information provided in Table 14.2-38, a preoperational test demonstrates that the chemical and volume control system ASME Code Class 3 check valves listed in Tier 1 Table 2.2-2 stroke fully closed under reverse flow conditions.

Preoperational test conditions are established that approximate design basis temperature, differential pressure and flow conditions Acceptance Criteria to the extent practicable, consistent with preoperational test limitations.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.02.05 CVCS The chemical and volume control system ASME Code Class 3 air- X operated valves are tested to demonstrate the capability to perform their function to fail to or maintain their position on loss of motive power under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-38, a preoperational test demonstrates that each chemical and volume control system ASME Code Class 3 air-operated valves listed in Tier 1 Table 2.2-2 fails closed on loss of motive power (electric power to the valve actuating solenoid(s) is lost, or pneumatic pressure to the valve(s) is lost).

Preoperational test conditions are established that approximate design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with 14.3-28 preoperational test limitations.

Certified Design Material and Inspections, Tests, Analyses, and 02.03.01 CES Section 5.2.5 Reactor Coolant Pressure Boundary Leakage X Detection, discusses that RCS leakage detection systems are designed to detect and, to the extent practicable, identify the source of reactor coolant leakage. The RCS leakage detection systems conform to the guidance of RG 1.45, regarding detection, monitoring, quantifying, and identification of reactor coolant leakage.

In accordance with the information provided in Table 14.2-41, a preoperational test demonstrates that the containment evacuation system (CES) detects a level increase in the CES sample vessel, which correlates to a detection of an unidentified RCS leakage rate of one gpm within one hour.

Acceptance Criteria Water vapor and non-condensable gases are removed from the containment vessel by the CES. The water vapor is collected and condensed in the CES sample vessel where it is monitored using level and temperature instrumentation. The CES sample vessel level instrumentation is used to quantify and trend leak rates in the Revision 1 containment.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.03.02 CES Section 5.2.5, Reactor Coolant Pressure Boundary Leakage X Detection, discusses that RCS leakage detection systems are designed to detect and, to the extent practicable, identify the source of reactor coolant leakage. The RCS leakage detection systems conform to the guidance of RG 1.45, regarding detection, monitoring, quantifying, and identification of reactor coolant leakage.

In accordance with Table 14.2-41, a preoperational test demonstrates that the CES is capable of detecting a pressure increase in the CES inlet pressure instrumentation (PIT-1001/PIT-1019), which correlates to a detection of an unidentified RCS leakage rate of one gpm within one hour.

02.04.01 TG Section 10.2.2.3.3, Overspeed Protection, provides a description of X the turbine generator system and its redundant independent 14.3-29 turbine overspeed protection systems (OSPs), i.e., the governor overspeed detection circuit and the turbine emergency trip system.

Certified Design Material and Inspections, Tests, Analyses, and An ITAAC inspection is performed of the turbine overspeed protection arrangement to verify that the trip circuitry for the governor overspeed detection circuit and the turbine emergency trip system are supplied from different power sources and do not share common equipment.

02.04.02 TG Section 10.1.2.4, Turbine Overspeed Protection, discusses the X turbine stop valve and turbine control valves and the associated turbine trip signals.

In accordance with the information provided in Table 14.2-33, a preoperational test will be performed to verify the turbine stop valve and turbine control valves close on a turbine overspeed trip Acceptance Criteria signal from both the turbine emergency trip system and the governor overspeed detection circuit.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.01 MPS Section 7.2.1.1, I&C Safety System Development Process, discusses X the software lifecycle phases for the MPS. The purpose is to verify software implementation based on licensing commitments to 10 CFR Part 50, Appendix A, GDC 1 (Quality), Appendix B (Quality Assurance Criteria), RGs 1.28, 1.152, 1.168, 1.169, 1.170, 1.171, 1.172, and 1.173, and the associated IEEE standards. The licensee shall perform analyses for each phase and generate technical reports to conclude that the lifecycle phases were implemented per the licensing commitments. Per RG 1.152, a generic waterfall software life cycle model consists of the following phases: (1) concepts, (2) requirements, (3) design, (4) implementation, (5) test, (6) installation, checkout, and acceptance testing, (7) operation, (8) maintenance, and (9) retirement.

The ITAAC verifies that output documentation of each Software Lifecycle phase satisfies the requirements of that phase for the MPS 14.3-30 and that software were implemented per licensing commitments Certified Design Material and Inspections, Tests, Analyses, and to 10 CFR Part 50, Appendix A, GDC1 (Quality), Appendix B (Quality Assurance Criteria), RGs 1.28, 1.152, 1.168, 1.169, 1.170, 1.171, 1.172, and 1.173, and the associated IEEE standards.

02.05.02 MPS Section 7.2.9, Control of Access, Identification, and Repair, X discusses the protective measures that prevent modification of the MPS tunable parameters without proper configuration and authorization. Guidance on this issue is provided in DI&C-ISG-04 Revision 1, "Highly-Integrated Control Rooms - Communications Issues," under interdivisional communications, staff position 10.

In accordance with Table 14.2-63, a preoperational test demonstrates that protective measures restrict modification to the Acceptance Criteria MPS tunable parameters without proper configuration and authorization. This test will be performed by attempting to modify the tunable parameters with the MPS not in the correct configuration or without authorization.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.03 MPS Sections 7.1.2, Independence, discusses the independence of the X MPS Class 1E I&C current-carrying circuits per the guidance of RG 1.75, which endorses IEEE Std. 384-1992. Physical separation is provided to maintain the independence of Class 1E I&C current-carrying circuits so that the safety functions required during and following any design basis event can be accomplished. Minimum separation distance (as defined in IEEE Std. 384-1992), or barriers or any combination thereof may achieve physical separation as specified in IEEE Std. 384-1992.

Separate ITAAC inspections are performed to verify the independence provided by physical separation and the independence provided by electrical isolation. This ITAAC verifies the independence of Class 1E current-carrying circuits by physical separation. An ITAAC inspection is performed of physical separation of the MPS Class 1E current-carrying circuits. The 14.3-31 physical separation ITAAC inspection results verify that the Certified Design Material and Inspections, Tests, Analyses, and following physical separation criteria are met:

i. Physical separation between redundant divisions of the MPS Class 1E I&C current-carrying circuits is provided by a minimum separation distance, or by barriers (where the minimum separation distances cannot be maintained), or by a combination of separation distance and barriers; and such physical separation satisfies the criteria of RG 1.75. The configuration of each as-built barrier agrees with its associated as-built drawing.

ii. Physical separation between the MPS Class 1E I&C current-carrying circuits and non-Class 1E I&C current-carrying circuits is provided by a minimum separation distance, or by barriers Acceptance Criteria (where the minimum separation distances cannot be maintained), or by a combination of separation distance and barriers; and such physical separation satisfies the criteria of RG 1.75. The configuration of each as-built barrier agrees with its associated as-built drawing.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.04 MPS Sections 7.1.2, Independence, discusses the independence of the X MPS Class 1E I&C circuits per the criteria of RG 1.75, which endorses IEEE Std. 384-1992. Electrical isolation is provided between the redundant divisions of the MPS Class 1E I&C circuits, and between Class 1E I&C circuits and non-Class 1E I&C circuits by Class 1E isolation devices so a failure in an I&C circuit does not prevent safety-related function completion in a different Class 1E I&C circuit.

An ITAAC inspection is performed to verify the following electrical isolation criteria are met:

i. Class 1E electrical isolation devices that satisfy the criteria of RG 1.75 are installed between redundant divisions of the MPS Class 1E I&C circuits.

ii. Class 1E electrical isolation devices that satisfy the criteria of 14.3-32 RG 1.75 are installed between the MPS Class 1E I&C circuits Certified Design Material and Inspections, Tests, Analyses, and and non-Class 1E I&C circuits.

02.05.05 MPS Sections 7.1.2, Independence, discusses the independence of MPS X Class 1E circuits per the criteria of RG 1.75, which endorses IEEE Std.

384-1992. Electrical isolation is provided between Class 1E circuits and non-Class 1E circuits by Class 1E isolation devices so a failure in a non-Class 1E circuit does not prevent the safety-related function completion in the Class 1E circuit.

i. The ITAAC verifies that: (1) an equipment qualification data report exists for the Class 1E isolation devices, and (2) the equipment qualification data report concludes that the Class 1E isolation devices performs its safety-related function under Acceptance Criteria the design basis environmental conditions specified in the equipment qualification data report.

ii. An ITAAC inspection is performed to verify that Class 1E electrical isolation devices are installed between MPS Class 1E circuits and non-Class 1E circuits, which satisfy the guidance Revision 1 of RG 1.75.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.06 MPS Section 7.1.2, Independence, discusses the communication X independence between redundant Class 1E digital communication system divisions. The purpose is to verify proper data isolation between redundant divisions. Requirements for independence are given in IEEE Std. 603-1991. Guidance for providing independence between redundant divisions of the Class 1E digital communication system is provided in Digital I&Cs Interim Staff Guidance (ISG) 04.

A vendor test demonstrates that independence between redundant divisions of the Class 1E MPS is provided.

02.05.07 MPS Section 7.1.2, Independence, discusses the communication X independence between Class 1E digital communication systems and non-Class 1E digital communication systems. The purpose is to verify that logical or software malfunction of the nonsafety-related system cannot affect the functions of the safety system.

14.3-33 Requirements for independence are given in IEEE Std. 603-1991.

Guidance for providing independence between the Class 1E digital Certified Design Material and Inspections, Tests, Analyses, and communication system and non-Class 1E digital communication systems is provided in Digital Instrumentation and Controls ISG 04.

A vendor test demonstrates that independence between the Class 1E MPS and non-Class 1E digital systems is provided.

02.05.08 MPS Section 7.1.1.2.1, Protection Systems, describes automatic and X manual reactor trips, variables that are monitored to provide input into automatic reactor trip signals, and the features of the reactor trip system (RTS). The reactor trip functions are listed in Table 7.1-3:

Reactor Trip Functions. The reactor trip logic for the monitored variables is provided in Figure 7.1-1.

The MPS initiates an automatic reactor trip signal when the Acceptance Criteria associated plant condition(s) exist.

In accordance with Table 14.2-63, a preoperational test demonstrates that a reactor trip signal is automatically initiated for each reactor trip function listed in Tier 1 Table 2.5-1.

Revision 1 The actuation of reactor trip breakers (RTBs) is not required for this test. The verification of the existence of a reactor trip signal is accomplished using main control room (MCR) displays.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.09 MPS Section 7.1.1.2.1, Protection Systems, describes automatic and X manual engineered safety features (ESFs) actuations, variables that are monitored to provide input into automatic ESFs signals, and the features of the ESF systems. The ESFs functions are listed in Table 7.1-4: Module Protection System Engineered Safeguards Functions.

The ESFs logic for the monitored variables is provided in Figure 7.1-1.

The MPS initiates an automatic ESF actuation signal when the associated plant condition(s) exist.

In accordance with Table 14.2-63, a preoperational test demonstrates that an automatic ESF actuation signal is automatically initiated for each of the ESF functions listed in Tier 1 Table 2.5-2.

The actuation of ESFs equipment is not required for this test. The 14.3-34 verification of the existence of an ESF actuation signal is Certified Design Material and Inspections, Tests, Analyses, and accomplished using MCR displays.

02.05.10 MPS Section 7.1.1.2.1, Protection Systems, describes automatic and X manual reactor trips, variables that are monitored to provide input into automatic reactor trip signals, and the features of the RTS. The reactor trip functions are listed in Table 7.1-3: Reactor Trip Functions. The reactor trip logic for the monitored variables is provided in Figure 7.1.

The MPS initiates an automatic reactor trip signal for the reactor trip functions when the associated plant condition(s) exist.

In accordance with Table 14.2-63, a preoperational test demonstrates that the RTBs open when any one of the automatic Acceptance Criteria reactor trip functions is initiated from the MCR. The RTBs are only opened once to satisfy this test objective.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.11 MPS Section 7.1.1.2.1, Protection Systems, describes automatic and X manual ESFs actuations, variables that are monitored to provide input into automatic ESFs signals, and the features of the engineered safety feature systems. The ESFs functions are listed in Table 7.1-4: Module Protection System Engineered Safeguards Functions. The ESFs logic for the monitored variables is provided in Figure 7.1.

The MPS initiates an automatic ESF actuation signal for the functions listed in Tier 1 Table 2.5-2 when the associated plant condition(s) exist.

In accordance with Table 14.2-63, a preoperational test demonstrates that ESF equipment automatically actuates to perform its safety-related function listed in Tier 1 Table 2.5-2 upon an injection of a single simulated MPS signal.

14.3-35 02.05.12 MPS Section 7.1.1.2.1, Protection Systems, describes automatic and X Certified Design Material and Inspections, Tests, Analyses, and manual reactor trips, variables that are monitored to provide input into automatic reactor trip signals, and the features of the RTS. A manual reactor trip is one of the MPS manually actuated functions.

In accordance with Table 14.2-63, a preoperational test demonstrates that the RTBs open when a reactor trip is manually initiated from the MCR.

02.05.13 MPS Section 7.1.1.2.1, Protection Systems, describes manual ESFs X actuation, variables that are monitored to provide input into automatic ESFs signals, and the features of the ESF system. The ESFs functions that can be manually actuated are shown in Figure 7.1-1 Acceptance Criteria In accordance with Table 14.2-63, a preoperational test demonstrates that the MPS actuates the ESF equipment to perform its safety-related function listed in Tier 1 Table 2.5-3 when manually initiated.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.14 MPS Section 7.1.6, Safety Evaluation, describes the MPS conformance to X the GDC in 10 CFR 50 Appendix A. Guidance provided in Design Specific Review Standard Section 7.2.3, Reliability, Integrity, and Completion of Protective Action, states that the design incorporate protective measures that provide for I&C safety systems to fail in a safe state, or into a state that has been demonstrated to be acceptable on some other defined basis, if conditions such as disconnection of the system, loss of power, or adverse environments, are experienced.

Section 7.1.6 describes that consistent with GDC 23, the MPS is designed, with sufficient functional diversity as to prevent the loss of a protection function, 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 power, or postulated adverse environments are experienced. Section 7.2.3.2, 14.3-36 System Integrity Characteristics, states that the MPS is designed Certified Design Material and Inspections, Tests, Analyses, and such that in the event of a condition such as a system disconnection or loss of power the MPS fails into a safe state.

In accordance with Table 14.2-63, a preoperational test demonstrates that when the loss of electrical power is detected in a separation group of the MPS that separation group fails to a safe state resulting in a reactor trip state for that separation group.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.15 MPS Section 7.1.6, Safety Evaluation, describes the MPS conformance to X the GDC in 10 CFR 50 Appendix A. Guidance provided in Design Specific Review Standard Section 7.2.3, Reliability, Integrity, and Completion of Protective Action, states that the design incorporate protective measures that provide for I&C safety systems to fail in a safe state, or into a state that has been demonstrated to be acceptable on some other defined basis, if conditions such as disconnection of the system, loss of power, or adverse environments, are experienced.

Section 7.1.6 describes that consistent with GDC 23, the MPS is designed, with sufficient functional diversity as to prevent the loss of a protection function, 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 power, or postulated adverse environments are experienced. Section 7.2.3.2, 14.3-37 System Integrity Characteristics, states that the MPS is designed Certified Design Material and Inspections, Tests, Analyses, and such that in the event of a condition such as a system disconnection or loss of power the MPS fails into a safe state. For an ESF function this predefined safe state may be that the actuated component remains as-is.

In accordance with Table 14.2-63, a preoperational test demonstrates that when the loss of electrical power is detected in a separation group of the MPS that separation group fails to a safe state for that separation group.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.16 MPS Section 7.2.3.3, Completion of Protective Action, describes X compliance with requirements for completion of protective actions, which requires that, once initiated, the reactor trip and ESF proceed to completion and remain in their required position/

condition until the actuation system is reset and operator action is taken. IEEE 603-1991 Clause 5.2 states that `The safety systems shall be designed so that, once initiated automatically or manually, the intended sequence of protective actions of the execute features shall continue until completion. Deliberate operation action shall be required to return the safety systems to normal. This requirement shall not preclude the use of equipment protective devices identified in [Clause] 4.11 of the design basis or the provisions for deliberate operator interventions. Seal-in of individual channels is not required."

In accordance with Table 14.2-63, a preoperational test 14.3-38 demonstrates that:

Certified Design Material and Inspections, Tests, Analyses, and

i. upon an MPS reactor trip signal listed in Tier 1 Table 2.5-1, the RTBs open and the RTBs do not automatically close when the MPS reactor trip signal clears.

ii. upon an MPS engineered safety feature actuation signal listed in Tier 1 Table 2.5-2, the ESF equipment actuates to perform its safety-related function and continues to maintain its safety-related position and perform its safety-related function when the MPS engineered safety feature actuation signal clears.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.17 MPS Section 7.2.12.1, Automatic Control, describes the signals and X initiating logic for each reactor trip and required response times.

Reactor trip response time is defined in technical specification Section 1.1, Definitions, as RTS RESPONSE TIME is that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

In accordance with Table 14.2-63, a preoperational test demonstrates that the measured time for the reactor trip functions listed in Tier 1 Table 2.5-1 is less than or equal to the maximum values assumed in the accident analysis. Technical specification Section 1.1, Definitions, states that the response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

Section 7.2.12.1, Automatic Control, describes the signals and 14.3-39 initiating logic for each ESF and the required response times. The ESF response time is defined in technical specification Section 1.1, Certified Design Material and Inspections, Tests, Analyses, and Definitions, as ESF RESPONSE TIME is that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions).

In accordance with Table 14.2-63, a preoperational test demonstrates that the measured time for the ESF functions listed in Tier 1 Table 2.5-2 is less than or equal to the maximum values assumed in the accident analysis. Technical specification Section 1.1, Definitions, states that the response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.18 MPS Section 7.2.4.1, Operating Bypasses, describes MPS operating X bypasses for reactor trip functions. Section 7.2.4.1, Operating Bypasses, describes MPS operating bypasses for ESF actuations.

The operating bypasses are applied automatically when plant conditions dictate that the safety function is not needed, or that the safety function prevents proper plant operation at a specific mode of operation.

In accordance with Table 14.2-63, a preoperational test demonstrates that the MPS interlocks listed in Tier 1 Table 2.5-4 automatically establish an operating bypass for the specified reactor trip or ESF actuations when a real or simulated signal simulates that the associated interlock condition is met; and are automatically removed when the real or simulated signal simulates that the associated permissive condition is no longer satisfied.

14.3-40 02.05.19 MPS Section 7.2.4.1, Operating Bypasses, describes MPS operating X bypasses for reactor trip functions. Section 7.2.4.1, Operating Certified Design Material and Inspections, Tests, Analyses, and Bypasses, describes MPS operating bypasses for ESF actuations.

The operating bypasses are applied automatically when plant conditions dictate that the safety function is not needed, or that the safety function prevents proper plant operation at a specific mode of operation.

In accordance with Table 14.2-63, a preoperational test demonstrates that the MPS permissives listed in Tier 1 Table 2.5-4 allows the manual bypass of the specified reactor trip or ESF actuations when a real or simulated signal simulates that the associated permissive condition is met; and are automatically removed when the real or simulated signal simulates that the Acceptance Criteria associated permissive condition is no longer satisfied.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.20 MPS Section 7.2.4.1, Operating Bypasses, describes MPS operating X bypasses for reactor trip functions. Section 7.2.4.1, Operating Bypasses, describes MPS operating bypasses for ESF actuations.

The operating bypasses are applied automatically when plant conditions dictate that the safety function is not needed, or that the safety function prevents proper plant operation at a specific mode of operation.

In accordance with Table 14.2-63, a preoperational test demonstrates that the MPS overrides listed in Tier 1 Table 2.5-4 are established when the manual override switch is active and a real or simulated RT-1 interlock is established.

02.05.21 MPS Section 7.2.4.2, Maintenance Bypass, describes the MPS X maintenance bypass operation mode. An individual protection channel can be placed in a maintenance bypass operation mode to 14.3-41 allow manual testing and maintenance during power operation, while ensuring that the minimum redundancy required by the Certified Design Material and Inspections, Tests, Analyses, and Technical Specifications is maintained. The reactor trip functions are listed in Table 7.1-3: Reactor Trip Functions. The ESFs functions are listed in Table 7.1-4: Module Protection System Engineered Safeguards Functions.

In accordance with Table 14.2-63, a preoperational test demonstrates that with a safety function module out of service switch activated, the safety function is placed in trip or bypass based on the position of the safety function module trip/bypass switch. Each separation group of the reactor trip functions listed in Tier 1 Table 2.5-1 and each separation group of the ESFs signals listed in Tier 1 Table 2.5-2 is tested by placing the separation group Acceptance Criteria in maintenance bypass.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.22 MPS Section 7.2.4.2, Maintenance Bypass, describes the MPS X maintenance bypass operation mode. An individual protection channel can be placed in a maintenance bypass operation mode to allow manual testing and maintenance during power operation, while ensuring that the minimum redundancy required by the technical specifications is maintained. Section 7.2.4.2 discusses the status indication of MPS manual or automatic bypasses placed in maintenance bypass operation mode.

In accordance with Table 14.2-63, a preoperational test demonstrates that each operational MPS manual or automatic bypass is indicated in the MCR.

02.05.23 MPS Section 7.2.4.2, Maintenance Bypass, describes the MPS X maintenance bypass operation mode. An individual protection channel can be placed in a maintenance bypass operation mode to 14.3-42 allow manual testing and maintenance during power operation, while ensuring that the minimum redundancy required by the Certified Design Material and Inspections, Tests, Analyses, and technical specifications is maintained. Section 7.2.4.2 discusses the status indication of MPS maintenance bypasses placed in maintenance bypass operation mode.

In accordance with Table 14.2-63, a preoperational test demonstrates that each MPS maintenance bypass is indicated in the MCR.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.24 MPS This ITAAC is intended to address self-testing features credited X towards surveillance or other operational testing. Given the nature of this ITAAC, it is acceptable to verify ITAAC completion during the factory acceptance testing (FAT). Self-testing features include, but are not limited to, watchdog timers, automated channel checks, and signal input comparisons.

Section 7.2.15.3, Fault Detection and Self-diagnostics, discusses the self-testing features of the MPS, including the types of faults that should be detected, the system responses to such faults, the required response times, and the ability for alarms and displays in the MCR to provide indication of such faults' existence.

These tests of the MPS self-testing features ensure that a) faults requiring detection are detected, b) the system responds appropriately to each fault based on the type of fault, c) the 14.3-43 response occurs within a sufficient timeframe to ensure safety function is not lost, and d) that alarms and indications in the main Certified Design Material and Inspections, Tests, Analyses, and control room indicate the type of fault present.

A vendor test demonstrates and a report exists and concludes that:

self-testing features verify that faults requiring detection are detected.

self-testing features verify that upon detection, the system responds according to the type of fault.

self-testing features verify that faults are detected and responded within a sufficient timeframe to ensure safety function is not lost.

self-testing features verify that detected faults are indicated by alarms and displays.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.25 MPS Section 7.1.1.2.2, Post Accident Monitoring, and Section 7.2.13, X Displays and Monitoring, describe the post-accident monitoring (PAM) Type B and C displays and alarms indicated on the safety display and indication system (SDIS) displays in the MCR. PAM Type B and C variables are developed in accordance with the guidance in RG 1.97, Revision 4, "Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants" which endorses (with certain clarifying regulatory positions specified in Section C of this guide) IEEE Std. 497-2002.1, "IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations."

In accordance with Table 14.2-66, a preoperational test demonstrates the ability to retrieve and display the various PAM Type B and C parameters and alarms at the as-built safety display indication displays in the main control room. The intent is to verify 14.3-44 that the displays and alarms function during testing of the Certified Design Material and Inspections, Tests, Analyses, and integrated as-built system; however, separate testing of the actual operation of the PAM alarms and displays using simulated signals may be acceptable where this is not practical.

02.05.26 MPS Section 18.6, Treatment of Important Human Actions, provides a X summary of the treatment of important human actions (TIHA) objectives, scope, methodology, and results. The TIHA methodology and the results are documented in the TIHA results summary report. The TIHA approach is consistent with the applicable provisions of NUREG-0711, Revision 3.

In accordance with Table 14.2-63, a preoperational test demonstrates that the minimum inventory of controls identified by Acceptance Criteria the human factors engineering process is can be manually operated from the operator workstation in the MCR.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.05.27 MPS Section 7.0.4.1.2, Reactor Trip System, discusses the arrangement of X the protection system RTBs. Figure 7.0-6: Reactor Trip Breaker Arrangement provides the arrangement of the RTBs.

This ITAAC verifies that the RTBs conform to the arrangement indicated in Tier 1 Figure 2.5-1. In addition, the ITAAC inspection verifies proper connection of the shunt and undervoltage trip mechanisms and other auxiliary contacts.

02.05.28 MPS Section 7.1.5.1, Application of NUREG/CR-6303 Guidelines, X discusses that two of the four separation groups and one of the two divisions of RTS and ESFAS will utilize a different programmable technology.

A ITAAC inspection is performed to verify that MPS separation groups A & C and Division I of RTS and ESFAS utilize a different programmable technology from separation groups B & D and 14.3-45 Division II of RTS and ESFAS.

Certified Design Material and Inspections, Tests, Analyses, and 02.05.29 MPS Section 7.1.3.3, Redundancy in Nonsafety I&C System Design, X discusses that when operators evacuate the MCR and occupy the RSS, two manual isolation switches for the MPS divisions are provided to isolate the MPS manual actuation switches in the MCR to prevent fires in the MCR from causing spurious actuations of associated equipment.

An ITAAC inspection is performed of each MCR isolation switch location to verify that the switch exists in the RSS.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.06.01 NMS Section 7.1.2.2, Electrical Independence, discusses the X independence of the neutron monitoring system (NMS) Class 1E circuits. Electrical isolation is provided between Class 1E circuits and non-Class 1E circuits by Class 1E isolation devices so a failure in a non-Class 1E circuit does not prevent the safety-related function completion in the Class 1E circuit.

A type test, analysis, or a combination of type test and analysis will be performed of the Class 1E isolation devices to verify that the Class 1E circuit does not degrade below defined acceptable operating levels when the non-Class 1E side of the isolation device is subjected to the maximum credible voltage, current transients, shorts, grounds, or open circuits.

An ITAAC inspection is performed to verify that Class 1E electrical isolation devices are installed between NMS Class 1E circuits and 14.3-46 non-Class 1E circuits.

Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.06.02 NMS Sections 7.0.3.2, Safety-Related Systems, 7.1.2.1, Physical X Independence, and 7.1.2.4.1, Independence between Redundant Portions of a Safety System, discuss the independence of the NMS Class 1E I&C current-carrying circuits per the guidance of RG 1.75, which endorses IEEE Std. 384-1992. Physical separation is provided to maintain the independence of Class 1E I&C current-carrying circuits so that the safety functions required during and following any design basis event can be accomplished. Minimum separation distance (as defined in IEEE Std. 384-1992), or barriers or any combination thereof may achieve physical separation as specified in IEEE Std. 384-1992.

Separate ITAAC inspections are performed to verify the independence provided by physical separation and the independence provided by electrical isolation. This ITAAC verifies the independence of Class 1E current-carrying circuits by physical 14.3-47 separation. An ITAAC inspection is performed of physical Certified Design Material and Inspections, Tests, Analyses, and separation of the NMS Class 1E current-carrying circuits. The physical separation ITAAC inspection results verify that the following physical separation criteria are met:

i. Physical separation between redundant divisions of the NMS Class 1E I&C current-carrying circuits is provided by a minimum separation distance, or by barriers (where the minimum separation distances cannot be maintained), or by a combination of separation distance and barriers; and such physical separation satisfies the criteria of RG 1.75. The configuration of each as-built barrier agrees with its associated as-built drawing.

Acceptance Criteria ii. Physical separation between the NMS Class 1E I&C current-carrying circuits and non-Class 1E I&C current-carrying circuits is provided by a minimum separation distance, or by barriers (where the minimum separation distances cannot be maintained), or by a combination of separation distance and Revision 1 barriers. The configuration of each as-built barrier agrees with its associated as-built drawing.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.06.03 NMS Sections 7.0.3.2, Safety-Related Systems, 7.1.2.2, Electrical X Independence, and 7.1.2.4.1, Independence between Redundant Portions of a Safety System, discuss the independence of the NMS Class 1E I&C circuits per the criteria of RG, which endorses IEEE Std.

384-1992. Electrical isolation is provided between the redundant divisions of the NMS Class 1E I&C circuits, and between Class 1E I&C circuits and non-Class 1E I&C circuits by Class 1E isolation devices so a failure in an I&C circuit does not prevent safety-related function completion in a different Class 1E I&C circuit.

An ITAAC inspection is performed to verify the following electrical isolation criteria are met:

i. Class 1E electrical isolation devices that satisfy the criteria of RG 1.75 are installed between redundant divisions of the NM system Class 1E I&C circuits.

14.3-48 ii. Class 1E electrical isolation devices that satisfy the criteria of RG Certified Design Material and Inspections, Tests, Analyses, and 1.75 are installed between the NMS Class 1E I&C circuits and non-Class 1E I&C circuits.

02.07.01 Section 11.5.2.2.7, Containment Evacuation System, discusses the X operation of the CES. For each high radiation signal listed in Tier 1 Table 2.7-1, the CES automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table.

In accordance with Table 14.2-41, a preoperational test demonstrates the CES automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table upon initiation of a real or simulated CES high radiation signal from CES-RT-1011.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.07.02 Section 11.5.2.2.11, Chemical and Volume Control System, X discusses the operation of the CVCS. For each high radiation signal listed in Tier 1 Table 2.7-1, the CVCS automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table.

In accordance with Table 14.2-38, a preoperational test demonstrates the CVCS automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table upon initiation of a real or simulated CVCS high radiation signal from CVC-RT-3016.

02.07.03 Section 11.5.2.2.14, Auxiliary Boiler System, discusses the operation X of the auxiliary boiler system (ABS) and the CVCS. For each high radiation signal listed in Tier 1 Table 2.7-1, the CVCS automatically aligns the components identified in Tier 1 Table 2.7-1 to the 14.3-49 required positions identified in the table.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with Table 14.2-38, a preoperational test demonstrates the CVCS automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table upon initiation of a real or simulated ABS high radiation signal from 6A-AB-RT-0142.

02.07.04 Section 11.5.2.2.14, Auxiliary Boiler System, discusses the operation X of the ABS and the CVCS. For each high radiation signal listed in Tier 1 Table 2.7-1, the CVCS automatically aligns the components identified in Tier 1 Table 2.7-1 to the required positions identified in the table.

In accordance with Table 14.2-38, a preoperational test demonstrates the CVCS automatically aligns the components Acceptance Criteria identified in Tier 1 Table 2.7-1 to the required positions identified in the table upon initiation of a real or simulated ABS high radiation signal from 6B-AB-RT-0141.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.01 EQ Section 3.10, Seismic and Dynamic Qualification of Mechanical and X Electrical Equipment, presents information to demonstrate that the Seismic Category I equipment, including its associated supports and anchorages, is qualified by type test, analysis, or a combination of type test and analysis to perform its function under the design basis seismic loads during and after an SSE. The qualification method employed for the Seismic Category I equipment is the same as the qualification method described for that type of equipment in Section 3.10. This method conforms to IEEE-344-2004 and ASME QME-1-2007 (or later editions), as accepted by the NRC staff in RG 1.100 Revision 3 (or later revision), with specific revision years and numbers as presented in Section 3.10.

The scope of equipment for this design commitment is module-specific, safety-related equipment, and module-specific, nonsafety-related equipment that has one of the following design 14.3-50 features:

Certified Design Material and Inspections, Tests, Analyses, and

Nonsafety-related mechanical and electrical equipment located within the boundaries of the NuScale Power Module that has an augmented Seismic Category I design requirement.

Nonsafety-related mechanical and electrical equipment that performs a credited function in Chapter 15 analyses (secondary main steam isolation valves (MSIV), feedwater regulating valves (FWRV) and secondary feedwater check valves.)

The ITAAC verifies that: (1) a seismic qualification record form exists for each Seismic Category I component type, and (2) the seismic qualification record form concludes that the Seismic Category I equipment listed in Tier 1 Table 2.8-1, including its associated Acceptance Criteria supports and anchorages, performs its function under the seismic design basis load conditions specified in the seismic qualification record form.

After installation in the plant, an ITAAC inspection is performed to verify that the Seismic Category I equipment listed in Tier 1 Table Revision 1 2.8-1, including its associated supports and anchorages, is installed in its design location in a Seismic Category I structure in a configuration bounded by the seismic qualification record form.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.02 EQ Section 3.11, Environmental Qualification of Mechanical and X Electrical Equipment, presents information to demonstrate that the electrical equipment, including its connection assemblies, located in a harsh environment is qualified by type test or a combination of type test and analysis to perform its function under design basis harsh environmental conditions, experienced during normal operations, anticipated operational occurrences, DBAs, and post-accident conditions in accordance with 10 CFR 50.49. As defined in IEEE-Std-572-2006, IEEE Standard for Qualification of Class 1E Connection Assemblies for Nuclear Power Generating Stations, a connection assembly is any connector or termination combined with related cables or wires as an assembly. The qualification method employed for the equipment is the same as the qualification method described for that type of equipment in Section 3.11.

14.3-51 The scope of equipment for this design commitment is Certified Design Material and Inspections, Tests, Analyses, and module-specific, Class 1E equipment located within a harsh environment, and module-specific, nonsafety-related equipment with an augmented equipment qualification design requirement located within the boundaries of the NuScale Power Module.

The ITAAC verifies that: (1) an equipment qualification record form exists for the electrical equipment listed in Tier 1 Table 2.8-1 and addresses connection assemblies, (2) the equipment qualification record form concludes that the electrical equipment, including its connection assemblies, performs its function under the environmental conditions specified in Section 3.11 and the equipment qualification record form, and (3) the required post-accident operability time for the electrical equipment in the Acceptance Criteria equipment qualification record form is in agreement with Section 3.11.

After installation in the plant, an ITAAC inspection is performed to verify that the electrical equipment listed in Tier 1 Table 2.8-1, Revision 1 including its connection assemblies, is installed in its design location in a configuration bounded by the equipment qualification record form.

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.03 EQ Section 3.11 presents information to demonstrate that the X non-metallic parts, materials, and lubricants used in safety-related mechanical equipment located in a harsh environment are qualified using a type test or a combination of type test and analysis to perform their function up to the end of their qualified life in design basis harsh environmental conditions experienced during normal operations, anticipated operational occurrences, DBAs, and post-accident conditions. Environmental conditions include both internal service conditions and external environmental conditions for the non-metallic parts, materials, and lubricant. The qualification method employed for the equipment is the same as the qualification method described for that type of equipment in Section 3.11.

The scope of equipment for this design commitment is module-specific, safety-related mechanical equipment, and 14.3-52 module-specific, nonsafety-related mechanical equipment that Certified Design Material and Inspections, Tests, Analyses, and performs a credited function in Chapter 15 analyses (secondary main steam isolation valves (MSIV), feedwater regulating valves (FWRV) and secondary feedwater check valves).

The ITAAC verifies that: (1) an equipment qualification record form or ASME QME-1 report exists for the non-metallic parts, materials, and lubricants used in mechanical equipment designated for a harsh environment, and (2) the qualification record form concludes that the non-metallic parts, materials, and lubricants used in mechanical equipment listed in Tier 1 Table 2.8-1 perform their intended function up to the end of its qualified life under the design basis environmental conditions (both internal service conditions and external environmental conditions) specified in the Acceptance Criteria qualification record form.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.04 EQ Section 3.11, Environmental Qualification of Mechanical and X Electrical Equipment, and Appendix 3C, Methodology for Environmental Qualification of Electrical and Mechanical Equipment, presents information to demonstrate that the Class 1E computer-based I&C systems located in a mild environment is qualified by type test or a combination of type test and analysis to perform its safety-related function under the design basis mild environmental conditions. The qualification method employed for the equipment is the same as the qualification method described for that type of equipment in Section 3.11 and Appendix 3C. This method conforms to IEEE-323-2003 (or later editions), as accepted by the NRC staff in RG 1.209, Revision 0 (or later revision), with specific revision years and numbers as presented in Section 3.10.

The ITAAC verifies that: (1) an equipment qualification record form exists for the Class 1E computer-based I&C systems listed in Tier 1 14.3-53 Table 2.8-1, and (2) the equipment qualification record form Certified Design Material and Inspections, Tests, Analyses, and concludes that the Class 1E computer-based I&C systems performs its safety-related function under the design basis mild environmental conditions specified in Section 3.11 and Appendix 3C and the equipment qualification record form.

After installation in the plant, an ITAAC inspection is performed to verify that the Class 1E computer-based I&C systems listed in Tier 1 Table 2.8-1 is installed in its design location in a configuration bounded by its equipment qualification record form.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.05 EQ Section 3.11, Environmental Qualification of Mechanical and X Electrical Equipment, presents information to demonstrate that the Class 1E digital equipment is qualified using a type test, analysis, or a combination of type test and analysis to perform its safety-related function when subjected to electromagnetic interference, radio frequency interference, and electrical surges that would exist before, during, and following a DBA. The qualification method employed for Class 1E digital equipment is the same as the qualification method described for that type of equipment in Section 3.11.

The ITAAC verifies that: (1) an equipment qualification record form exists for the Class 1E digital equipment listed in Tier 1 Table 2.8-1, and (2) the equipment qualification record form concludes that the Class 1E digital equipment withstands the design basis electromagnetic interference, radio frequency interference, and 14.3-54 electrical surges that would exist before, during, and following a Certified Design Material and Inspections, Tests, Analyses, and DBA without loss of safety-related function.

02.08.06 EQ Section 3.9.6.1, Functional Design and Qualification of Pumps, X Valves, and Dynamic Restraints, and Section 3.10.2, Methods and Procedures for Qualifying Mechanical and Electrical Equipment and Instrumentation, discuss that the functional qualification of safety-related valves is performed in accordance with ASME QME 2007(or later edition), as accepted in RG 1.100 Revision 3 (or later revision), with specific revision years and numbers as presented in Section 3.9.6.1. The qualification method employed for the valves agrees with the qualification method described in Section 3.10.2.

The ITAAC verifies that: (1) A Functional Qualification Report exists Acceptance Criteria for the safety-related valves listed in Tier 1 Table 2.8-1, and (2) the Functional Qualification Report concludes that safety-related valves are capable of performing their safety-related function under the full range of fluid flow, differential pressure, electrical conditions, and temperature conditions up to and including DBA conditions.

Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.07 EQ Section 3.9.3.2, Design and Installation of Pressure Relief Devices, X discusses that relief valves provide overpressure protection in accordance with the ASME Code Section III.

The ITAAC verifies that: (1) the test for each relief valve listed in Tier 1 Table 2.8-1 meets the set pressure, capacity, and overpressure design requirements; and (2) each relief valve listed in Tier 1 Table 2.8-1 is provided with an ASME Code Certification Mark that identifies the valve's set pressure, capacity, and overpressure.

02.08.08 EQ Section 5.4.2, Decay Heat Removal System, discusses that the DHRS X passive condensers provide the safety-related function of transferring their design heat load from the DHRS during shutdown. After manufacture of the DHRS passive condensers, a type test or a combination of type test and analysis is performed to validate that the DHRS passive condensers are capable of meeting 14.3-55 the specified heat transfer performance requirements. Section 5.4.2 discusses the design heat transfer capabilityof the DHR system Certified Design Material and Inspections, Tests, Analyses, and passive condensers.

The ITAAC verifies that the safety-related passive condensers listed in Tier 1 Table 2.8-1 have a heat removal capacity sufficient to transfer their design heat load.

Acceptance Criteria Revision 1

Table 14.3-1: Module-Specific Structures, Systems, and Components Based Design Features Tier 2 NuScale Final Safety Analysis Report and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 02.08.09 EQ Section 3.11, Environmental Qualification of Mechanical and X Electrical Equipment, presents information to demonstrate that the CNTS electrical penetration assemblies, including its connection assemblies, located in a harsh environment are qualified by type test or a combination of type test and analysis to perform its safety-related function under design basis harsh environmental conditions, experienced during normal operations, anticipated operational occurrences, DBAs, and post-accident conditions in accordance with 10 CFR 50.49. As defined in IEEE-Std-572-2006, IEEE Standard for Qualification of Class 1E Connection Assemblies for Nuclear Power Generating Stations, a connection assembly is any connector or termination combined with related cables or wires as an assembly. The qualification method employed for the equipment is the same as the qualification method described for that type of equipment in Section 3.11.

14.3-56 The ITAAC verifies that: (1) an equipment qualification record form Certified Design Material and Inspections, Tests, Analyses, and exists for the CNTS electrical penetration assemblies listed in Tier 1 Table 2.8-1and addresses connection assemblies; (2) the equipment qualification record form concludes that the CNTS electrical penetration assemblies, including its connection assemblies, performs its safety-related function under the environmental conditions specified in Section 3.11 and the equipment qualification record form; and (3) the required post-accident operability time for the CNTS electrical penetration assemblies in the equipment qualification record form is in agreement with Section 3.11.

After installation in the plant, an ITAAC inspection is performed to verify that the CNTS electrical penetration assemblies listed in Tier Acceptance Criteria 1 Table 2.8-1, including its connection assemblies, is installed in its design location in a configuration bounded by the equipment qualification record form.

Revision 1

Tier 2 NuScale Final Safety Analysis Report Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.01.01 CRH Testing is performed on the CRE in accordance with RG 1.197, X Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, to demonstrate that air exfiltration from the CRE is controlled. RG 1.197 allows two options for CRE testing; either integrated testing (tracer gas testing) or component testing. Section 6.4 Control Room Habitability, describes the testing requirements for the CRE habitability program. Section 6.4 provides the maximum air exfiltration allowed from the CRE.

In accordance with Table 14.2-18, a preoperational test using the tracer gas test method demonstrates that the air exfiltration from the CRE does not exceed the assumed unfiltered leakage rate provided in Table 6.4-1: Control Room Habitability System Design 14.3-57 Parameters for the dose analysis. Tracer gas testing in accordance with ASTM E741 will be performed to measure the unfiltered in-Certified Design Material and Inspections, Tests, Analyses, and leakage into the CRE with the control room habitability system (CRHS) operating.

03.01.02 CRH The CRHS valves are tested by remote operation to demonstrate X the capability to perform their function to transfer open and transfer closed under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-18, a preoperational test demonstrates that each CRHS valve listed in Tier 1 Table 3.1-1 strokes fully open and fully closed by remote operation under preoperational test conditions.

Preoperational test conditions are established that approximate Acceptance Criteria design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.01.03 CRH The CRHS solenoid-operated valves are tested to demonstrate the X capability to perform their function to fail open on loss of motive power under preoperational temperature, differential pressure, and flow conditions.

In accordance with Table 14.2-18 a preoperational test demonstrates that each CRHS solenoid-operated valve listed in Tier 1 Table 3.1-1 repositions to the open position on loss of motive power (electric power to the valve actuating solenoid(s) is lost, or pneumatic pressure to the valve(s) is lost).

Preoperational test conditions are established that approximate design basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

03.01.04 CRH Section 6.4.4, Design Evaluation, discusses the thermal mass of the X 14.3-58 CRB and its contents limit the temperature increase as shown in Certified Design Material and Inspections, Tests, Analyses, and Table 6.4-3 within the CRE within an acceptable range for the first 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following a DBA.

An analysis confirms that the CRE bulk average air temperature is acceptable on a loss of active cooling for the first 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following a DBA.

03.01.05 CRH Section 6.4.3.2, Off-Normal Operation, discusses the operation of X the CRHS, which maintains a positive pressure in the MCR relative to the adjacent areas. Table 6.4-1: Control Room Habitability System Design Parameters provides the required positive pressure in the MCR relative to the adjacent areas. In accordance with Table 14.2-18, a preoperational test demonstrates that the CRHS maintains a positive pressure of greater than or equal to 1/8 Acceptance Criteria inches water gauge in the MCR relative to adjacent areas, while operating in a DBA alignment.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.02.01 CRV The normal control room HVAC system (CRVS) control room X X envelope isolation dampers are tested to demonstrate the capability to perform their function to fail to the closed position on loss of motive power.

In accordance with Table 14.2-19, a preoperational test demonstrates that each CRVS air-operated CRE isolation damper listed in Tier 1 Table 3.2-1 repositions to the closed position on loss of motive power (electric power to the valve actuating solenoid is lost, or pneumatic pressure to the damper is lost).

Preoperational test conditions are established that approximate design differential pressure conditions to the extent practical, consistent with preoperational test limitations. A manual signal, actual automatic signal, or simulated automatic signal may be used to operate the valves because the control logic of the valves 14.3-59 is not being verified by this ITAAC.

Certified Design Material and Inspections, Tests, Analyses, and 03.02.02 CRV Section 9.4.1.2, System Description, discusses the operation of the X CRVS, which maintains a positive pressure in the CRB relative to the outside environment. This is consistent with the requirements of 10 CFR Part 20, Subparts E and H and 10 CFR Part 50, Appendix I.

In accordance with Table 14.2-19 a preoperational test demonstrates that the CRVS will maintain a positive pressure of greater than or equal to 1/8 inches water gauge in the CRB relative to the outside environment, while operating in a normal operating alignment.

03.02.03 CRV Section 9.4.1.2.2.1, Normal Operation, provides a discussion of X how the CRVS maintains the hydrogen concentration levels in the CRB battery rooms containing batteries below one percent by Acceptance Criteria volume.

In accordance with Table 14.2-19, a preoperational test demonstrates that the airflow capability of the CRVS maintains the hydrogen concentration levels in the CRB battery rooms Revision 1 containing batteries below one percent by volume.

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.03.01 RBV Section 9.4.2.2.2.1, Normal Operation, and Section 9.4.2.2.2.2, Off- X normal Operation, discuss the operation of the Reactor Building HVAC system (RBVS) which maintains a negative pressure in the RXB relative to the outside environment. This is consistent with the requirements of 10 CFR Part 20, Subparts E and H and 10 CFR Part 50, Appendix I.

In accordance with Table 14.2-20, a preoperational test demonstrates that the RBVS will maintain a negative pressure in the RXB relative to the outside environment, while operating in a normal operating alignment.

03.03.02 RBV Section 9.4.2.2.2.1, Normal Operation, and Section 9.4.2.2.2.2, Off- X normal Operation, discuss the operation of the RBVS which maintains a negative pressure in the RWB relative to the outside environment. This is consistent with the requirements of 10 CFR 14.3-60 Part 20, Subparts E and H and 10 CFR Part 50, Appendix I.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with Table 14.2-20, a preoperational test demonstrates that the RBVS will maintain a negative pressure in the RWB relative to the outside environment, while operating in a normal operating alignment.

03.03.03 RBV Section 9.4.2.2.2.1, Normal Operation, provides a discussion of X how the RBVS maintains the hydrogen concentration levels in the RXB battery rooms containing batteries below one percent by volume.

In accordance with Table 14.2-20, a preoperational test demonstrates that the airflow capability of the RBVS maintains the hydrogen concentration levels in the RXB battery rooms containing batteries below one percent by volume.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.04.01 FHE Section 9.1.4, Fuel Handling Equipment, describes that the fuel X handling machine (FHM) is classified as a Type I crane as defined by the ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).

An ITAAC inspection is performed of the FHM main and auxiliary hoists machinery arrangement to verify the existence of the following single-failure proof features: (a) nonredundant structural components (bridge, trolley, wire rope drum, and hook) are designed to appropriate standards, constructed from base material demonstrated to meet appropriate material properties, and, (b) redundant design features to stop and hold the load following:

specified component failures (e.g., wire rope, drive train, and control system) 14.3-61

operator errors (e.g., two-blocking and overload)

Certified Design Material and Inspections, Tests, Analyses, and This ITAAC inspection may be performed any time after manufacture of the FHM (at the factory or later).

03.04.02 FHE Section 9.1.4, Fuel Handling Equipment, describes that the FHM is X classified as a Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder). The FHM main hoist is tested in accordance with the applicable requirements of NOG-1. An FAT demonstrates that the single failure proof FHM main hoist is rated load tested at 125% (+5%, -0%) of the manufacturers rating.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.04.03 FHE Section 9.1.4, Light-Load Handling System (Related to Refueling), X describes that the FHM is classified as a Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder). In accordance with ASME NOG-1 paragraph 7422, the FHM auxiliary hoist is full load tested at a maximum of 100% of the hoist manufacturer's rating. After the full load test is completed, and prior to use of the crane to handle loads, the FHM auxiliary hoist is rated load tested at 125% (+5%, -0%) of the manufacturer's rating in accordance with ASME NOG-1 paragraph 7423.

An FAT demonstrates that the single failure proof FHM auxiliary hoist is rated load tested at 125% (+5%, -0%) of the manufacturer's rating.

03.04.04 FHE Section 9.1.4, Light-Load Handling System (Related to Refueling), X 14.3-62 describes that the single failure proof FHM is classified as a Type I crane as defined by the ASME NOG-1 Code, Rules for Construction Certified Design Material and Inspections, Tests, Analyses, and of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).

An ITAAC inspection is performed to verify that the ASME Type I as-built FHM welds are nondestructively examined in accordance with the standards of ASME NOG-1 paragraph 4251.4 and the FHM purchase specification.

This ITAAC inspection may be performed any time after manufacture of the single failure-proof FHM (at the factory or later).

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.04.05 FHE Section 9.1.4, Light-Load Handling System (Related to Refueling), X provide descriptions of how the limit switches on the FHM gripper mast limits travel and that the fuel handling equipment (FHE) has provisions to limit maximum lift height of a fuel assembly to maintain a water inventory of 10 feet above the top of the fuel assembly for personnel shielding.

In accordance with Table 14.2-51, a preoperational test demonstrates that the FHM maintains at least 10 feet of water above the top of the fuel assembly when lifted to its maximum height with the pool level at the lower limit of the normal operating low water level.

03.04.06 FHE Section 9.1.4, Light-Load Handling System (Related to Refueling), X provides description of how the new fuel jib crane hook movement is limited to prevent carrying a fuel assembly over the 14.3-63 spent fuel pool to prevent a load drop onto the spent fuel racks.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with Table 14.2-51, a preoperational test demonstrates that the new fuel jib crane interlocks prevent the crane from carrying a fuel assembly over the spent fuel racks.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.05.01 SFS The ASME Code Section III requires that documentary evidence be X available at the construction or installation site before use or installation to ensure that ASME Code Class NF components conform to the requirements of the Code. The Fuel Storage system ASME Code Class NF components require a Data Report as specified by NCA-1210. The Data Report is prepared by the certificate holder or owner and signed by the certificate holder or owner and the Inspector as specified by NCA-8410. The type of individual Data Report Forms necessary to record the required Code Data is specified in Table NCA-8100-1.

An ITAAC inspection is performed of the Data Reports for Fuel Storage system ASME Code Class NF as-built fuel storage racks that are described in Section 9.1.2 to (1) ensure that the appropriate Data Reports have been provided as specified in Table NCA-8100-1 and (2) ensure that the certificate holder or owner 14.3-64 and the authorized nuclear inspector have signed the Data Certified Design Material and Inspections, Tests, Analyses, and Reports, and (3) verify that the requirements of ASME Code Section III are met.

03.05.02 SFS Section 9.1.1, Criticality Safety of Fresh and Spent Fuel Storage and X Handling, discusses the criticality analysis of the fuel storage racks.

An ITAAC inspection is performed to verify that the as-built fuel storage racks, including any neutron absorbers, conform to the design values for materials and dimensions and their tolerances, as presented in the approved criticality analysis.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.06.01 UHS As required by ASME Code Section III NCA-1210, each ASME Code X Class 1, 2 and 3 component (including piping systems) of a nuclear power plant requires a Design Report in accordance with NCA-3550. NCA-3551.1 requires that the drawings used for construction be in agreement with the Design Report before it is certified and be identified and described in the Design Report. It is the responsibility of the N certificate holder to furnish a Design Report for each component and support, except as provided in NCA-3551.2 and NCA-3551.3. NCA-3551.1 also requires that the Design Report be certified by a registered professional engineer when it is for Class 1 components and supports, Class CS core support structures, Class MC vessels and supports, Class 2 vessels designed to NC-3200 (NC-3131.1), or Class 2 or Class 3 components designed to service loadings greater than design loadings. NCA-3554 requires that any modification of any 14.3-65 document used for construction, from the corresponding document used for design analysis, shall be reconciled with the Certified Design Material and Inspections, Tests, Analyses, and Design Report.

An ITAAC inspection is performed of the ultimate heat sink (UHS)

ASME Code Class 3 as-built piping system Design Report to verify that the e requirements of ASME Code Section III are met.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.06.02 UHS Section 9.1.2.2.2, Spent Fuel Storage, Section 9.1.3.2.1, Spent Fuel X Pool Cooling System, and Section 9.1.3.2.2, Reactor Pool Cooling System, discuss spent fuel pool (SFP) and reactor pool cooling. No piping, openings, doors, or penetrations through the SFP, refueling pool, reactor pool and dry dock walls are installed below the minimum water level required for shielding, spent fuel cooling, DHRS cooling, or containment cooling. Gates, openings, or drains, permanently connected mechanical or hydraulic systems (piping), and other features that by maloperation or failure could reduce the coolant inventory to unsafe levels are not included in the design.

An ITAAC inspection is performed to verify that the SFP, refueling pool, reactor pool and dry dock include no drains, piping or other systems below 80 ft building elevation (55 ft pool level as measured from the bottom of the SFP and reactor pool). This 14.3-66 inspection is performed by physical measurements in the as-built Certified Design Material and Inspections, Tests, Analyses, and SFP and reactor pool.

03.07.01 FP Section 9.5.1.2.6, Fire Protection Design Features, discusses how X the fire protection system (FPS) meets the guidance provided by RG 1.189 and applicable NFPA standards. Two separate dedicated 100 percent capacity freshwater storage tanks are provided.

An ITAAC inspection is performed to verify that the minimum usable water volume of each firewater storage tank is greater than or equal to 300,000 gallons. If the storage tanks are also used as backup water sources for other non-fire emergencies, the ITAAC inspection verifies that the non-fire emergencies cannot drain the tank below the minimum dedicated useable water volume of Acceptance Criteria 300,000 gallons required for firefighting.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.07.02 FP Section 9.5.1, Fire Protection Program, discusses how the capacity X of each FPS pump is adequate to supply the total flow demand at the pressure required at the pump discharge. Section 9.5.1 provides the design flow of the fire pumps.

i. An analysis confirms that the as-built fire pumps provide the flow demand for the largest sprinkler or deluge system plus an additional 500 gpm for fire hoses assuming failure of the largest fire pump or loss of off-site power.

ii. In accordance with Table 14.2-25, a preoperational test demonstrates that each fire pump delivers the design flow to the FPS while operating in the fire-fighting alignment.

14.3-67 Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.07.03 FP Section 9.5.1 discusses that (a) safe-shutdown can be achieved X assuming that all equipment in any one fire area (except for the MCR and containment) is rendered inoperable by fire and that reentry into the fire area for repairs and operator actions is not possible, (b) that smoke, hot gases, or fire suppressant cannot migrate from the affected fire area into other fire areas to the extent that they could adversely affect safe shutdown capabilities, including operator actions, and (c) an independent alternative shutdown capability that is physically and electrically independent of the MCR exists.

A safe shutdown analysis of the as-built plant will be performed, including a post-fire safe shutdown circuit analysis performed in accordance with RG 1.189 and NEI 00-01for all possible fire-induced failures that could affect the safe shutdown success path, including multiple spurious actuations.

14.3-68 The safe shutdown analysis will verify that:

Certified Design Material and Inspections, Tests, Analyses, and

safe shutdown can be achieved assuming that all equipment in any one fire area (except for the MCR and containment) is rendered inoperable by fire and that reentry into the fire area for repairs and operator actions is not possible.

smoke, hot gases, or fire suppressant cannot migrate from the affected fire area into other fire areas to the extent that they could adversely affect safe shutdown capabilities, including operator actions.

an independent alternative shutdown capability that is physically and electrically independent of the MCR exists.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.07.04 FP Appendix 9A, Fire Hazards Analysis, discusses the methodology X and presents the fire hazards analysis (FHA) for each fire area. The FHA must reflect the as-built configuration of the plant. The FHA is an analysis of the fire hazards, including combustible loading and ignition sources, and analysis of the fire protection features required to mitigate each postulated fire.

An FHA of the as-built plant will be performed in accordance with RG 1.189, as described in Appendix 9A. The FHA will verify (1) combustible loads and ignition sources are accounted for, and (2) fire protection features are suitable for the hazards they are intended for.

03.08.01 PL Section 9.5.3, Lighting Systems, discusses the plant lighting X X system (PLS) which provides normal illumination of the operator workstations and SDIS panels in the MCR and operator 14.3-69 workstations in the RSS. The PLS is capable of delivering at least 100 foot-candles of illumination to the MCR seated operator Certified Design Material and Inspections, Tests, Analyses, and stations and 50 foot-candles of illumination to the MCR primary operating areas and remote and auxiliary operating panels. Lower illumination levels may be used within these areas to ensure more favorable visual conditions, or for areas where critical tasks are not performed.

In accordance with Table 14.2-60, a preoperational test demonstrates that the PLS provides at least:

i. 100 foot-candles illumination at the MCR operator workstations and at least 50 foot-candles at the MCR auxiliary panels.

Acceptance Criteria ii. 100 foot-candles illumination at the RSS operator workstations.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.08.02 PL Section 9.5.3 discusses the PLS which provides emergency X X illumination of the operator workstations and SDIS panels in the MCR and operator workstations in the RSS.

In accordance with Table 14.2-60, a preoperational test demonstrates that the PLS provides at least:

i. 10 foot-candles of illumination at the MCR operator workstations and MCR auxiliary panels.

ii. 10 foot-candles at the RSS operator workstations.

03.08.03 PL Section 9.5.3 discusses the use of eight-hour battery pack X X emergency lighting fixtures, which provide illumination of at least one foot-candle for post-fire safe shutdown activities outside of the MCR and RSS. These units should provide lighting for:

areas required for power restoration / recovery to comply with 14.3-70 the guidance of RG 1.189, Fire Protection for Nuclear Power Certified Design Material and Inspections, Tests, Analyses, and Plants.

areas where normal actions are required for operation of equipment needed during fire; and

escape or access routes for firefighting and the remote shutdown area.

In accordance with the requirements in Table 14.2-60, a preoperational test demonstrates that eight-hour battery pack emergency lighting fixtures illuminate their required target areas to provide at least one foot-candle illumination in the areas outside the MCR or RSS where post-fire safe-shutdown activities are performed.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.09.01 RM Section 11.5.2.2.1, Normal Control Room HVAC System, discusses X the operation of the CRVS. For each high radiation signal listed in Tier 1 Table 3.9-1, the CRVS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

In accordance with Table 14.2-19, a preoperational test demonstrates the CRVS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated CRVS high radiation signal from 00-CRV-RT-0503, 00-CRV-RT-0504, and 00-CRV-RT-0505.

03.09.02 RM Section 11.5.2.2.1, Normal Control Room HVAC System, and X Section 11.5.2.2.2, Control Room Habitability System, discuss the operation of the CRVS and CRHS. For each high radiation signal 14.3-71 listed in Tier 1 Table 3.9-1, the CRVS and the CRHS automatically align the components identified in Tier 1 Table 3.9-1 to the Certified Design Material and Inspections, Tests, Analyses, and required positions identified in the table.

In accordance with Table 14.2-18, a preoperational test demonstrates the CRVS and the CRHS automatically align the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated CRVS high radiation signal from 00-CRV-RT-0510 and 00-CRV-RT-0511.

03.09.03 RM Section 11.5.2.2.3, Reactor Building HVAC System, discusses the X operation of the RBVS. For each high radiation signal listed in Tier 1 Table 3.9-1, the RBVS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified Acceptance Criteria in the table.

In accordance with Table 14.2-20, a preoperational test demonstrates the RBVS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated RBVS high Revision 1 radiation signal from 00-RBV-RE-0510, 00-RBV-RE-0511, and 00-RBV-RE-0512.

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.09.04 RM Section 11.5.2.2.6, Gaseous Radioactive Waste System, discusses X the operation of the gaseous radioactive waste system (GRWS).

For each high radiation signal listed in Tier 1 Table 3.9-1, the GRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

In accordance with Table 14.2-36, a preoperational test demonstrates the GRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated GRWS high radiation signal from 00-GRW-RIT-0046.

03.09.05 RM Section 11.5.2.2.6, Gaseous Radioactive Waste System, discusses X the operation of the GRWS. For each high radiation signal listed in Tier 1 Table 3.9-1, the GRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified 14.3-72 in the table.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with Table 14.2-36, a preoperational test demonstrates the GRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated GRWS high radiation signal from 00-GRW-RIT-0060.

03.09.06 RM Section 11.5.2.2.6, Gaseous Radioactive Waste System, discusses X the operation of the GRWS. For each high radiation signal listed in Tier 1 Table 3.9-1, the GRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

In accordance with Table 14.2-36, a preoperational test demonstrates the GRWS automatically aligns the components Acceptance Criteria identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated GRWS high radiation signal from 00-GRW-RIT-0071.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.09.07 RM Section 11.5.2.1.5, Liquid Radioactive Waste System, discusses the X operation of the liquid radioactive waste system (LRWS). For each high radiation signal listed in Tier 1 Table 3.9-1, the LRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

In accordance with Table 14.2-35, a preoperational test demonstrates the LRWS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated LRWS high radiation signal from 00-LRW-RIT-0569 and 00-LRW-RIT-0571.

03.09.08 RM Section 11.5.2.2.14, Auxiliary Boiler System, discusses the X operation of the ABS. For each high radiation signal listed in Tier 1 Table 3.9-1, the ABS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified 14.3-73 in the table.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with Table 14.2-9, a preoperational test demonstrates the ABS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated ABS high radiation signal from 00-AB-RT-0153.

03.09.09 RM Section 11.5.2.2.14, Auxiliary Boiler System, discusses the X operation of the ABS. For each high radiation signal listed in Tier 1 Table 3.9-1, the ABS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

In accordance with Table 14.2-9, a preoperational test demonstrates the ABS automatically aligns the components Acceptance Criteria identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated ABS high radiation signal from 00-AB-RT-0166.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.09.10 RM Section 11.5.2.1.4, Pool Surge Control System, discusses the X operation of the pool surge control system (PSCS). For each high radiation signal listed in Tier 1 Table 3.9-1, the PSCS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table.

In accordance with Table 14.2-4, a preoperational test demonstrates the PSCS automatically aligns the components identified in Tier 1 Table 3.9-1 to the required positions identified in the table upon initiation of a real or simulated PSCS high radiation signal from 00-PSC-RE-1003.

03.10.01 RBC Section 9.1.5, Overhead Heavy Load Handling System, describes X that the Reactor Building crane (RBC) main hoist is classified as a Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top Running 14.3-74 Bridge, Multiple Girder).

Certified Design Material and Inspections, Tests, Analyses, and An ITAAC inspection is performed of the RBC main hoist machinery arrangement to verify the existence of the following single-failure proof features: (a) nonredundant structural components (bridge, trolley, wire rope drum, and hook) are designed to appropriate standards, constructed from base material demonstrated to meet appropriate material properties, and, (b) redundant design features to stop and hold the load following:

specified component failures (e.g., wire rope, drive train, and control system)

operator errors (e.g., two-blocking and overload)

Acceptance Criteria This ITAAC inspection may be performed any time after manufacture of the RBC (at the factory or later).

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.10.02 RBC Section 9.1.5 describes that the RBC auxiliary hoists are classified X as a Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).

An ITAAC inspection is performed of the RBC auxiliary hoists machinery arrangement to verify the existence of the following single-failure proof features: (a) nonredundant structural components (wire rope drum, and hook) are designed to appropriate standards, constructed from base material demonstrated to meet appropriate material properties, and, (b) redundant design features to stop and hold the load following:

specified component failures (e.g., wire rope, drive train, and control system)

operator errors (e.g., two-blocking and overload) 14.3-75 This ITAAC inspection may be performed any time after Certified Design Material and Inspections, Tests, Analyses, and manufacture of the RBC auxiliary hoists (at the factory or later).

03.10.03 RBC Section 9.1.5 describes that the RBC wet hoist is classified as a X Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).

An ITAAC inspection is performed of the RBC wet hoist arrangement to verify the existence of the following single-failure proof features: (a) nonredundant structural components (bridge, trolley, wire rope drum, and hook) are designed to appropriate standards, constructed from base material demonstrated to meet appropriate material properties, and, (b) redundant design Acceptance Criteria features to stop and hold the load following:

specified component failures (e.g., wire rope, drive train, and control system)

operator errors (e.g., two-blocking and overload)

Revision 1 This ITAAC inspection may be performed any time after manufacture of the RBC wet hoist (at the factory or later).

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.10.04 RBC Section 9.1.5 describes that the RBC main hoist is classified as a X Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder). In accordance with ASME NOG-1, the RBC main hoist is full load tested at a maximum of 100% of the hoist manufacturers rating. After the full load test is completed, and prior to use of the crane to handle loads, the RBC is rated load tested at 125% (+5%, -0%) of the manufacturers rating in accordance with ASME NOG-1, paragraph 7423.

An FAT demonstrates that the single failure-proof RBC main hoist is rated load tested at 125% (+5%, -0%) of the manufacturers rating.

03.10.05 RBC Section 9.1.5 describes that the RBC auxiliary hoists are classified X as a Type I crane as defined by the ASME NOG-1 Code, Rules for 14.3-76 Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder). In accordance with ASME NOG-1, each Certified Design Material and Inspections, Tests, Analyses, and RBC auxiliary hoist is full load tested at a maximum of 100% of the hoist manufacturers rating. After the full load test is completed, and prior to use of the RBC auxiliary hoists to handle loads, each RBC auxiliary hoist is rated load tested at 125% (+5%, -0%) of the manufacturers rating in accordance with ASME NOG-1, paragraph 7423.

An FAT demonstrates that the single failure proof RBC auxiliary hoists are rated load tested at 125% (+5%, -0%) of the manufacturers rating.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.10.06 RBC Section 9.1.5 describes that the RBC wet hoist is classified as a X Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder). In accordance with ASME NOG-1, the RBC wet hoist is full load tested at a maximum of 100% of the hoist manufacturers rating. After the full load test is completed, and prior to use of the RBC wet hoist to handle loads, the RBC wet hoist is rated load tested at 125% (+5%, -0%) of the manufacturers rating in accordance with ASME NOG-1, paragraph 7423.

An FAT demonstrates that the single failure proof RBC wet hoist is rated load tested at 125% (+5%, -0%) of the manufacturers rating.

03.10.07 RBC Section 9.1.5 discusses that the single failure-proof RBC is X classified as a Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top 14.3-77 Running Bridge, Multiple Girder).

Certified Design Material and Inspections, Tests, Analyses, and An ITAAC inspection is performed to verify that the ASME Type I as-built RBC welds are nondestructively examined in accordance with the standards of ASME NOG-1.

This ITAAC inspection may be performed any time after manufacture of the single failure proof RBC (at the factory or later).

03.10.08 RBC Section 9.1.5 discusses that the single failure-proof RBC wet hoist X is classified as a Type I crane as defined by the ASME NOG-1 Code, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder).

An ITAAC inspection is performed to verify that the ASME Type I as-built RBC wet hoist welds are nondestructively examined in Acceptance Criteria accordance with the standards of ASME NOG-1.

This ITAAC inspection may be performed any time after manufacture of the single failure-proof RBC wet hoist (at the factory or later).

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.11.01 RXB Section 9.5.1, Fire Protection Program, discusses that fire and X X smoke barriers separate: (1) safety-related systems from any potential fires in nonsafety-related areas that could affect the ability of safety-related systems to perform their safety-related function; (2) redundant divisions or trains of safety-related systems from each other to prevent damage that could adversely affect a safe shutdown function from a single fire; (3) equipment within a single safety-related electrical division that present a fire hazard to equipment in another safety-related division; (4) electrical circuits (safety-related and nonsafety-related) whose fire-induced failure could cause a spurious actuation that could adversely affect a safe shutdown function.

An ITAAC inspection is performed to verify that the following RXB as-built fire barriers and smoke barriers are installed in accordance with the FHA and are qualified for the fire rating specified in the 14.3-78 FHA:

Certified Design Material and Inspections, Tests, Analyses, and

fire-rated doors

fire-rated penetration seals

fire-rated dampers

smoke barriers

fire-rated walls, floors, and ceilings The objective of the inspection is to verify that the fire and smoke barriers meet the design requirements, location requirements, and that they are qualified for their intended use based upon visual inspection and review of the as-built drawings and Acceptance Criteria qualification documentation.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.11.02 RXB Section 3.4.1, Internal Flood Protection for Onsite Equipment X Failures, discusses the features used to mitigate the consequences of internal flooding, which include structural enclosures, barriers, curbs, sills, and watertight seals.

An ITAAC inspection is performed to verify that the following RXB as-built internal flooding barriers are installed in accordance with the internal flooding analysis report and are qualified as specified in the internal flooding analysis report:

flood resistant doors

curbs and sills

walls

water tight penetration seals 14.3-79

NEMA enclosures Certified Design Material and Inspections, Tests, Analyses, and The objective of the inspection is to verify that the flooding barriers meet the design requirements, location requirements, and that they are qualified for their intended use based upon visual inspection and review of the as-built drawings and qualification documentation.

03.11.03 RXB Section 2.4.2, Floods, discusses that the maximum flood elevation X (including wind-induced wave run-up) is one foot below baseline plant elevation. Section 3.4.2.1, Probable Maximum Flood, states that the probable maximum flood elevation (including wave action) of the design is one foot below the baseline plant elevation (100'-0).

Acceptance Criteria An ITAAC inspection is performed to verify that the RXB as-built floor elevation at ground entrances is located above the maximum external flood elevation to protect the RXB from external flooding. The inspection will compare the maximum external flood elevation against the RXB as-built design drawings Revision 1 to verify that the floor elevation at ground entrances is a minimum of one foot above the maximum external flood elevation.

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.11.04 RXB Section 12.3, Radiation Protection Design Features, provides the X design bases for radiation shielding, including type, form and material properties utilized in specific locations. Radiation shielding is provided to meet the radiation zone and access requirements for normal operation and post-accident conditions, and to demonstrate compliance with 10 CFR 50.49, GDC 4, and GDC 19. Compartment walls, ceilings, and floors, or other barriers provide shielding.

An ITAAC inspection is performed to verify that the thickness of RXB radiation barriers is greater than or equal to the required thicknesses. The required thicknesses are specified in Tier 1 Table 3.11-1.

03.11.05 RXB Section 12.3.2.2, Design Considerations, provides the design bases X for radiation shielding. Radiation shielding is provided to meet the 14.3-80 radiation zone requirements for normal operation and control room access requirements for post-accident conditions. Radiation Certified Design Material and Inspections, Tests, Analyses, and attenuating doors must meet or exceed the radiation attenuation capability of the wall within which they are installed.

An ITAAC inspection is performed to verify that the RXB radiation attenuating doors are installed in their design location and have a radiation attenuation capability that meets or exceeds that of the wall within which they are installed in accordance with the approved door schedule design.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.11.06 RXB Section 3.8.4.4, Design and Analysis Procedures, and Appendix 3B, X NuScale Plant Critical Sections, provide descriptive information, including plans and sections of each Seismic Category I structure, to establish that there is sufficient information to define the primary structural aspects and elements relied upon for the structure to perform the intended safety functions. Critical dimensions are identified in Appendix 3B. The RXB and its design basis loads are discussed in Section 3.8.4.3, Loads and Load Combinations. Critical sections are the subcomponents of individual Seismic Category I structures (i.e., shear walls, floor slabs and roofs, structure-to-structure connections) that are analytically representative of an essentially complete design. Design basis load combinations are shown in Table 3.8.4-3 and Table 3.8.4-4 and may include:

D = Dead loads, including piping, equipment, and partitions.

14.3-81

F = Loads due to weight and pressures of fluids.

Certified Design Material and Inspections, Tests, Analyses, and

H = Loads due to weight and static pressure of soil, water in soil, or other materials.

L = Live loads due to occupancy and moveable equipment.

Lr = Roof live load.

Ro = Piping and equipment reaction loads.

Ra = Piping and equipment reaction loads due to a postulated accident.

To = Thermal loads due to normal operating temperatures.

Acceptance Criteria

Ta = Thermal loads due to accident condition temperatures.

R = Rain load.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident

S = Snow load.

Se = Extreme snow load.

W = Straight line wind load.

Wt = Loads due to the design basis tornado.

Wh = Loads due to the design basis hurricane.

Eo = Seismic load due to an operating basis earthquake.

Ess = Seismic load due to an SSE.

Ccr = Loads due to the RBC.

Pa = Pressure loads due to accident conditions.

Yj = Jet impingement load generated by a postulated pipe break.

14.3-82

Yr = Loads on the structure generated by the reaction of the Certified Design Material and Inspections, Tests, Analyses, and broken pipe during a postulated break.

Ym = Missile impact load, or related internal moments and forces, on the structure generated by a postulated pipe break.

B = Loads due to buoyant force.

Guidance for the content and structure of the design report is provided in Standard Review Plan Section 3.8.4, Appendix C as shown in Table 3.B-2.

An ITAAC inspection and analysis is performed to ensure that deviations between the drawings used for construction and of the as-built RXB are reconciled and the RXB maintains its structural Acceptance Criteria integrity under the design basis loads. The design report provides criteria for the reconciliation between design and as-built conditions.

An ITAAC inspection is performed of the as-built RXB to verify that Revision 1 the dimensions of the RXB critical sections listed in Appendix B, Table 3B-54, conform to the approved design.

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.11.07 RXB Section 3.2.1, Seismic Classification, discusses that per RG 1.29, X some SSC that perform no safety-related functions could, if they failed under seismic loading, prevent or reduce the functioning of Seismic Category I SSC.

An ITAAC inspection and analysis is performed to verify that the as-built non-Seismic Category I SSC where a potential for adverse interaction with a Seismic Category I SSC exists will not impair the ability of Seismic Category I SSC to perform their safety functions as demonstrated by one or more of the following criteria:

Seismic Category I SSC are isolated from non-Seismic Category I SSC so that interaction does not occur.

Seismic Category I SSC are analyzed to confirm that the ability to perform their safety functions is not impaired as a result of impact from non-Seismic Category I SSC.

14.3-83

A non-Seismic Category I restraint system designed to Seismic Certified Design Material and Inspections, Tests, Analyses, and Category I requirements is used to assure that no interaction occurs between Seismic Category I SSC and non-Seismic Category I SSC.

03.11.08 RXB Section 3.6, Protection against Dynamic Effects Associated with X X Postulated Rupture of Piping, provides the design bases and criteria for the analysis required to demonstrate that safety-related SSC are not impacted by the adverse effects of a high-and moderate-energy pipe failure within the plant.

An ITAAC inspection is performed to verify that the as-built protective features located in the RXB outside the reactor pool bay credited in the reconciled Pipe Break Hazards Analysis Report Acceptance Criteria (such as pipe whip restraints, pipe whip or jet impingement barriers, jet impingement shields, or guard pipe) have been installed in accordance with design drawings of sufficient detail to show the existence and location of the protective hardware. The as-built inspection is intended to verify that changes to postulated Revision 1 pipe failure locations and protective features or protected equipment made during construction do not adversely affect the safety-related functions of the protected equipment.

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.12.01 RWB Section 12.3, Radiation Protection Design Features, provides the X design bases for radiation shielding, including type, form and material properties utilized in specific locations. Radiation shielding is provided to meet the radiation zone requirement for normal operation and post-accident conditions and to demonstrate conformance with RG 4.21 and RG 8.8. Compartment walls, ceilings, and floors, or other barriers provide shielding.

An ITAAC inspection is performed to verify that the thickness of RWB radiation barriers is greater than or equal to the required thicknesses. The required thicknesses are specified in Tier 1 Table 3.12-1.

03.12.02 RWB Section 12.3.2.2, Design Considerations, provides the design bases X for radiation shielding. Radiation shielding is provided to meet the radiation zone requirements for normal operation and post-14.3-84 accident conditions, and to demonstrate conformance to RG 4.21 and RG 8.8. Radiation attenuating doors must meet or exceed the Certified Design Material and Inspections, Tests, Analyses, and radiation attenuation capability of the wall within which they are installed.

An ITAAC inspection is performed to verify that the RWB radiation attenuating doors are installed in their design location and have a radiation attenuation capability that meets or exceeds that of the wall within which they are installed in accordance with the approved door schedule design.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.12.03 RWB The RW-IIa RWB and its design basis loads are discussed in Section X X 3.8.4.1.3, Radioactive Waste Building. Guidance for the content and structure of the as-built design report is provided in Standard Review Plan Section 3.8.4, Appendix C.

The scope of this ITAAC is a reconciliation of deviations between the issued for construction drawings that implement the seismic and dynamic analyses and the as-built structures. The design report provides criteria for the reconciliation. Design basis loads for RW-IIa structures as listed in RG 1.143 are:

earthquake

wind

tornado 14.3-85

tornado missile Certified Design Material and Inspections, Tests, Analyses, and

flood

precipitation (rain, snow)

accidental explosion (fixed facility)

accidental explosion (transportation vehicle)

malevolent vehicle assault

small aircraft crash An ITAAC inspection and reconciliation analysis is performed of the as-built RW-IIa RWB to ensure that deviations between the drawings used for construction and the as-built RW-IIa RWB are Acceptance Criteria reconciled and the as-built RW-IIa RWB maintains its structural integrity under the design basis loads. The design report provides criteria for the reconciliation between design and as-built conditions.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.13.01 CRB Section 9.5.1, Fire Protection Program, discusses that fire and X X smoke barriers separate: (1) Safety-related systems from any potential fires in nonsafety-related areas that could affect the ability of safety-related systems to perform their safety-related function. (2) Redundant divisions or trains of safety-related systems from each other to prevent damage that could adversely affect a safe shutdown function from a single fire. (3) Equipment within a single safety-related electrical division that present a fire hazard to equipment in another safety-related division. (4)

Electrical circuits (safety-related and nonsafety-related) whose fire-induced failure could cause a spurious actuation that could adversely affect a safe shutdown function.

An ITAAC inspection is performed to verify that the following CRB as-built fire barriers and smoke barriers are installed in accordance with the FHA and are qualified for the fire rating specified in the 14.3-86 FHA:

Certified Design Material and Inspections, Tests, Analyses, and

fire-rated doors

fire-rated penetration seals

fire-rated dampers

smoke barriers The objective of the inspection is to verify that the fire and smoke barriers meet the design requirements, location requirements, and that they are qualified for their intended use based upon visual inspection and review of the as-built drawings and qualification documentation.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.13.02 CRB Section 3.4.1, Internal Flood Protection for Onsite Equipment X Failures, discusses the features used to mitigate the consequences of internal flooding, which include structural enclosures, barriers, and watertight seals.

An ITAAC inspection is performed to verify that the following CRB as-built internal flooding barriers are installed in accordance with the internal flooding analysis report and are qualified as specified in the internal flooding analysis report:

flood resistant doors

walls

water tight penetration seals

NEMA enclosures 14.3-87 The objective of the inspection is to verify that the flooding Certified Design Material and Inspections, Tests, Analyses, and barriers meet the design requirements, location requirements, and that they are qualified for their intended use based upon visual inspection and review of the as-built drawings and qualification documentation.

03.13.03 CRB Section 2.4.2, Floods, discusses that the maximum flood elevation X (including wind-induced wave run-up) is one foot below baseline plant elevation. Section 3.4.2.1, Probable Maximum Flood, states that the probable maximum flood elevation (including wave action) of the design is one foot below the baseline plant elevation (100'-0).

An ITAAC inspection is performed to verify that the CRB as-built floor elevation at ground entrances is located above the Acceptance Criteria maximum external flood elevation to protect the CRB from external flooding. The inspection will compare the maximum external flood elevation against the CRB as-built design drawings to verify that the floor elevation at ground entrances is a minimum of one foot above the maximum external flood elevation.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.13.04 CRB Section 3.8.4.4, Design and Analysis Procedures, and Appendix 3B, X NuScale Plant Critical Sections, provide descriptive information, including plans and sections of each Seismic Category I structure, to establish that there is sufficient information to define the primary structural aspects and elements relied upon for the structure to perform the intended safety functions. Critical dimensions are identified in Appendix 3B. The CRB at Elevation 120'-0" and below and its design basis loads are discussed in Section 3.8.4.3, Loads and Load Combinations. Critical sections are the subcomponents of individual Seismic Category I structures (i.e., shear walls, floor slabs and roofs, structure-to-structure connections) that are analytically representative of an essentially complete design. Design basis loads load combinations are shown in Table 3.8.4-3 and Table 3.8.4-4 and may include:

14.3-88 Certified Design Material and Inspections, Tests, Analyses, and Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident

D = Dead loads, including piping, equipment, and partitions.

F = Loads due to weight and pressures of fluids.

H = Loads due to weight and static pressure of soil, water in soil, or other materials.

L = Live loads due to occupancy and moveable equipment.

Lr = Roof live load.

Ro = Piping and equipment reaction loads.

Ra = Piping and equipment reaction loads due to a postulated accident.

To = Thermal loads due to normal operating temperatures.

Ta = Thermal loads due to accident condition temperatures.

14.3-89

R = Rain load.

Certified Design Material and Inspections, Tests, Analyses, and

S = Snow load.

Se = Extreme snow load.

W = Straight line wind load.

Wt = Loads due to the design basis tornado.

Wh = Loads due to the design basis hurricane.

Ess = Seismic load due to an SSE.

Pa = Pressure loads due to accident conditions.

Acceptance Criteria Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident Guidance for the content and structure of the design report is provided in Standard Review Plan Section 3.8.4, Appendix C as shown in Table 3.B-2.

An ITAAC inspection and analysis is performed to ensure that deviations between the drawings used for construction and of the as-built CRB are reconciled. The design report provides criteria for the reconciliation between design and as-built conditions.

An ITAAC inspection is performed of the as-built CRB at Elevation 120-0 and below to verify that the dimensions of the CRB critical sections listed in Appendix B, Table 3B-55, conform to the approved design.

03.13.05 CRB Section 3.2.1, Seismic Classification, discusses that per RG 1.29, X some SSC that perform no safety-related functions could, if they failed under seismic loading, prevent or reduce the functioning of 14.3-90 Seismic Category I SSC.

Certified Design Material and Inspections, Tests, Analyses, and An ITAAC inspection and analysis is performed to verify that the as-built non-Seismic Category I SSC located where a potential for adverse interaction with a Seismic Category I SSC exists will not impair the ability of Seismic Category I SSC to perform their safety functions as demonstrated by one or more of the following criteria:

The collapse of the non-Seismic Category I structure will not cause the non-Seismic Category I structure to strike a Seismic Category I SSC.

The collapse of the non-Category I structure will not impair the Acceptance Criteria integrity of Seismic Category I SSC, nor result in incapacitating injury to control room occupants.

The non-Category I structure will be analyzed and designed to prevent its failure under SSE conditions.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.14.01 EQ Section 3.10, Seismic and Dynamic Qualification of Mechanical X and Electrical Equipment, presents information to demonstrate that the Seismic Category I equipment, including its associated supports and anchorages, is qualified by type test, analysis, or a combination of type test and analysis to perform its function under the design basis seismic loads during and after an SSE. The qualification method employed for the Seismic Category I equipment is the same as the qualification method described for that type of equipment in Section 3.10. This method conforms to IEEE-344-2004 and ASME QME-1-2007 (or later editions), as accepted by the NRC staff in RG 1.100 Revision 3 (or later revision),

with specific revision years and numbers as presented in Section 3.10.

The scope of equipment for this design commitment is the common, safety-related equipment, and the common, 14.3-91 nonsafety-related equipment that provides one of the following Certified Design Material and Inspections, Tests, Analyses, and nonsafety-related functions:

Provides physical support of irradiated fuel (fuel handling machine, spent fuel storage racks, reactor building crane, and module lifting adapter)

Provides a path for makeup water to the UHS

Provides containment of UHS water

Monitors UHS water level The ITAAC verifies that: (1) a seismic qualification record form exists for each Seismic Category I component type, and (2) the seismic qualification record form concludes that the Seismic Category I equipment listed in Tier 1 Table 3.14-1, including its Acceptance Criteria associated supports and anchorages, performs its function under the seismic design basis load conditions specified in the seismic qualification record form.

After installation in the plant, an ITAAC inspection is performed to verify that the Seismic Category I equipment listed in Tier 1 Table Revision 1 3.14-1, including its associated supports and anchorages, is installed in its design location in a Seismic Category I structure in a configuration bounded by the seismic qualification record form.

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.14.02 EQ Section 3.11, Environmental Qualification of Mechanical and X Electrical Equipment, presents information to demonstrate that the common electrical equipment, including its connection assemblies, located in a harsh environment is qualified by type test or a combination of type test and analysis to perform its function under design basis harsh environmental conditions, experienced during normal operations, anticipated operational occurrences, DBAs, and post-accident conditions in accordance with 10 CFR 50.49. As defined in IEEE-Std-572-2006, IEEE Standard for Qualification of Class 1E Connection Assemblies for Nuclear Power Generating Stations, a connection assembly is any connector or termination combined with related cables or wires as an assembly. The qualification method employed for the equipment is the same as the qualification method described for that type of equipment in Section 3.11.

14.3-92 The scope of equipment for this design commitment is the Certified Design Material and Inspections, Tests, Analyses, and nonsafety-related equipment that provides monitoring of the UHS water level and the non-safety related electrical equipment on the fuel handling machine and reactor building crane used to physically support irradiated fuel.

The ITAAC verifies that: (1) an equipment qualification record form exists for the common electrical equipment listed in Tier 1 Table 3.14-1 and addresses connection assemblies, (2) the equipment qualification record form concludes that the common electrical equipment, including its connection assemblies, performs its function under the environmental conditions specified in Section 3.11 and the equipment qualification record form, and (3) the required post-accident operability time for the common electrical Acceptance Criteria equipment in the equipment qualification record form is in agreement with Section 3.11.

After installation in the plant, an ITAAC inspection is performed to verify that the common electrical equipment listed in Tier 1 Table Revision 1 3.14-1, including its connection assemblies, is installed in its design location in a configuration bounded by the equipment qualification record form.

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.15.01 HFE Section 18.11, Design Implementation, describes the implementation of HFE aspects of the plant design.

An ITAAC inspection is performed to verify that the as-built configuration of main control room HSI is consistent with the as-designed configuration of main control room HSI as modified by the Integrated System Validation Report.

03.16.01 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC. Vital 14.3-93 equipment and vital area are discussed in the report.

Certified Design Material and Inspections, Tests, Analyses, and An ITAAC inspection is performed of the as built vital equipment to verify that the equipment is located in a vital area.

03.16.02 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC.

Provisions for accessing vital equipment are discussed in the report.

Acceptance Criteria An ITAAC inspection is performed of the as built vital equipment location to verify that access to vital equipment, within the nuclear island and structures, requires passage through at least two physical barriers, or as otherwise identified in Technical Report TR-0416-48929, "NuScale Design of Physical Security Revision 1 Systems."

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.16.03 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess,impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC.

A type test, analysis, or a combination of type test and analysis are performed of the bullet-resisting barriers used in the external walls, doors, ceilings and floors in the MCR, central alarm station, and the last access control function for access to the protected area. This qualification will demonstrate that the barriers are bullet-resistant, to Underwriters Laboratories Ballistic Standard 752, "The Standard of Safety for Bullet-Resisting Equipment," Level 14.3-94 4, or National Institute of Justice Standard 0108.01, "Ballistic Certified Design Material and Inspections, Tests, Analyses, and Resistant Protective Materials," Type III.

03.16.04 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC. The access control system which limits access to vital areas, within the nuclear island and structures, to individuals with unescorted access authorization is discussed in the report.

Acceptance Criteria In accordance with Table 14.2-73, a preoperational test demonstrates that the access control system provides authorized access to vital areas, within the nuclear island and structures, only to those individuals with authorization for unescorted access.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.16.05 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC. The report discusses that unoccupied vital area portals, within the nuclear island and structures, are equipped with locking devices and alarms that annunciate in the central alarm station.

In accordance with Table 14.2-74, a preoperational test, inspection, or a combination of test and inspection demonstrates that unoccupied vital areas, within the nuclear island and structures, are locked and alarmed and intrusion is detected and 14.3-95 annunciated in the central alarm station as described in Technical Certified Design Material and Inspections, Tests, Analyses, and Report TR-0416-48929, "NuScale Design of Physical Security Systems."

03.16.06 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security related information that describes security design bases and requirements for security SSC. The central alarm station and their location are discussed in the report.

Acceptance Criteria An ITAAC inspection is performed of the as built central alarm station to verify that it is located inside the protected area and the interior is not visible from the protected area perimeter.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.16.07 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC. Security alarms, within the nuclear island and structures, are discussed in the report.

In accordance with Table 14.2-74, a preoperational test demonstrates that:

(1) alarm annunciation indicates the type of alarm and location.

(2) security alarm devices, including transmission lines to 14.3-96 annunciators, are tamper-indicating and self-checking.

Certified Design Material and Inspections, Tests, Analyses, and (3) an automatic indication is provided when failure of the alarm system or a component occurs or when the system is on standby power.

03.16.08 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC. The Acceptance Criteria intrusion detection and assessment system, within the nuclear island and structures, is discussed in the report.

In accordance with Table 14.2-74, a preoperational test demonstrates that the intrusion detection and assessment system, within the nuclear island and structures, provides visual and Revision 1 audible annunciation of alarms in the central alarm station.

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.16.09 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC. The intrusion detection and assessment system, within the nuclear island and structures, is discussed in the report.

In accordance with Table 14.2-74, a preoperational test demonstrates that the intrusion detection and assessment system, within the nuclear island and structures, records each onsite security alarm annunciation, including each alarm, false alarm, alarm check, and tamper indication that identifies the type of 14.3-97 alarm, location, alarm circuit, date, and time.

Certified Design Material and Inspections, Tests, Analyses, and 03.16.10 Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC.

Emergency exits vital area boundaries, within the nuclear island and structures, are discussed in the report.

In accordance with Table 14.2-74, a preoperational test, Acceptance Criteria inspection, or a combination of test and inspection demonstrates that emergency exits through the vital area boundaries, within the nuclear island and structures, are alarmed with intrusion detection devices and secured by locking devices that allow egress during an emergency as described in Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems."

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.16.11 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC. The central alarm station's landline telephone service is discussed in the report.

In accordance with Table 14.2-68, a preoperational test, inspection, or a combination of test and inspection demonstrates that the central alarm station is equipped with conventional landline telephone service with the MCR and with local law enforcement authorities as described in Technical Report TR-0416-14.3-98 48929, "NuScale Design of Physical Security Systems."

Certified Design Material and Inspections, Tests, Analyses, and 03.16.12 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC. The central alarm station's communication system is discussed in the report.

In accordance with Table 14.2-68, a preoperational test, Acceptance Criteria inspection, or a combination of test and inspection demonstrates that the central alarm station is capable of continuous communication with on-duty security force personnel as described in Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems."

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.16.13 SEC Section 13.6 discusses that the physical security system design provides the capabilities to detect, assess, impede, and delay threats up to and including the design basis threat, and to provide for defense-in-depth through the integration of systems, technologies, and equipment. Technical Report TR-0416-48929, "NuScale Design of Physical Security Systems," provides safeguards and security-related information that describes security design bases and requirements for security SSC.

Nonportable communications equipment in the central alarm station is discussed in the report.

In accordance with Table 14.2-68, a preoperational test, inspection, or a combination of test and inspection demonstrates that nonportable communications equipment in the central alarm station remains operable (without disruption) from an independent power source in the event of loss of normal power as 14.3-99 described in Technical Report TR-0416-48929, "NuScale Design of Certified Design Material and Inspections, Tests, Analyses, and Physical Security Systems."

03.17.01 RM Section 11.5.2.2.9, Containment Flooding and Drain System, X discusses the operation of the containment flooding and drain system (CFDS). For each high radiation signal listed in Tier 1 Table 3.17-1, the CFDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-42, a preoperational test demonstrates the CFDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table upon initiation of a real or Acceptance Criteria simulated CFDS high radiation signal from 6A-CFD-RT-1007.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.17.02 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the balance-of-plant drain system (BPDS). For each high radiation signal listed in Tier 1 Table 3.17-1, the BPDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-24, a preoperational test demonstrates the BPDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high radiation signal from 6A-BPD-RIT-0552.

03.17.03 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the BPDS. For each high radiation signal listed in Tier 1 Table 3.17-1, the BPDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified 14.3-100 in the table.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with the information presented in Table 14.2-24, a preoperational test demonstrates the BPDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high radiation signal from 6A-BPD-RIT-0529.

03.17.04 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the BPDS. For each high radiation signal listed in Tier 1 Table 3.17-1, the BPDS automatically aligns the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-24, a preoperational test demonstrates the BPDS automatically aligns Acceptance Criteria the components identified in Tier 1 Table 3.17-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high radiation signal from 6A-BPD-RIT-0705.

Revision 1

Table 14.3-2: Shared/Common Structures, Systems, and Components and Non-Structures, Systems, and components Based Tier 2 NuScale Final Safety Analysis Report Design Features and Inspections, Tests, Analyses, and Acceptance Criteria Cross Reference (Continued)

ITAAC No. System Discussion DBA Internal/External Radiological PRA & Severe FP Hazard Accident 03.18.01 RM Section 11.5.2.2.9, Containment Flooding and Drain System, X discusses the operation of the containment flooding and drain system (CFDS). For each high radiation signal listed in Tier 1 Table 3.18-1, the CFDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-42, a preoperational test demonstrates the CFDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table upon initiation of a real or simulated CFDS high radiation signal from 6B-CFD-RT-1007.

03.18.02 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the BPDS. For each high radiation signal listed in Tier 1 Table 3.18-1, the BPDS automatically aligns the components 14.3-101 identified in Tier 1 Table 3.18-1 to the required positions identified in the table.

Certified Design Material and Inspections, Tests, Analyses, and In accordance with the information presented in Table 14.2-24, a preoperational test demonstrates the BPDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high radiation signal from 6B-BPD-RIT-0552.

03.18.03 RM Section 11.5.2.2.15, Balance-of-Plant Drain System, discusses the X operation of the BPDS. For each high radiation signal listed in Tier 1 Table 3.18-1, the BPDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table.

In accordance with the information presented in Table 14.2-24, a Acceptance Criteria preoperational test demonstrates the BPDS automatically aligns the components identified in Tier 1 Table 3.18-1 to the required positions identified in the table upon initiation of a real or simulated BPDS high radiation signal from 6B-BPD-RIT-0529.

Revision 1

ML18086A185

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