Text

LLC Submittal of the NuScale Standard Design Approval Application Part 2 - Final Safety Analysis Report, Chapter 14, Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria, Revision 0
	ML22362A070
Person / Time
Site:	99902078, 05200050
Issue date:	12/28/2022
From:	Fosaaen C; NuScale
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	LO-131959
	Download: ML22362A070 (1)
	v • d • e

LO-131959 December 28, 2022 Docket No.52-050 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of the NuScale Standard Design Approval Application Part 2 - Final Safety Analysis Report, Chapter 14, Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria, Revision 0

REFERENCES:

1. NuScale letter to NRC, NuScale Power, LLC Submittal of Planned Standard Design Approval Application Content, dated February 24, 2020 (ML20055E565)

2. NuScale letter to NRC, NuScale Power, LLC Requests the NRC staff to conduct a pre-application readiness assessment of the draft, NuScale Standard Design Approval Application (SDAA), dated May 25, 2022 (ML22145A460)

3. NRC letter to NuScale, Preapplication Readiness Assessment Report of the NuScale Power, LLC Standard Design Approval Draft Application, Office of Nuclear Reactor Regulation dated November 15, 2022 (ML22305A518)

4. NuScale letter to NRC, NuScale Power, LLC Staged Submittal of Planned Standard Design Approval Application, dated November 21, 2022 (ML22325A349)

NuScale Power, LLC (NuScale) is pleased to submit Chapter 14 of the Standard Design Approval Application, Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria, Revision 0. This chapter supports Part 2, Final Safety Analysis Report, (FSAR) of the NuScale Standard Design Approval Application (SDAA) (Reference 1). NuScale submits the chapter in accordance with requirements of 10 CFR 52 Subpart E, Standard Design Approvals. As described in Reference 4, the enclosure is part of a staged SDAA submittal.

NuScale requests NRC review, approval, and granting of standard design approval for the US460 standard plant design.

From July 25, 2022 to October 26, 2022, the NRC performed a pre-application readiness assessment of available portions of the draft NuScale FSAR to determine the FSARs readiness for submittal and for subsequent review by NRC staff (References 2 and 3). The NRC staff reviewed draft Chapter 14. The NRC did not identify readiness issues with the chapter. contains SDAA Part 2 Chapter 14, Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria, Revision 0.

This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.

NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-131959 Page 2 of 2 12/28/2022 If you have any questions, please contact Mark Shaver at 541-360-0630 or at mshaver@nuscalepower.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December 28, 2022.

Sincerely, Carrie Fosaaen Senior Director, Regulatory Affairs NuScale Power, LLC Distribution: Brian Smith, NRC Michael Dudek, NRC Getachew Tesfaye, NRC Bruce Bavol, NRC David Drucker, NRC Enclosure 1: SDAA Part 2 Chapter 14, Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-131959 :

SDAA Part 2 Chapter 14, Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria, Revision 0 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale US460 Plant Standard Design Approval Application Chapter Fourteen Initial Test Program and Inspections, Tests, Analyses, and Acceptance Criteria Final Safety Analysis Report Revision 0

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. . . . . . . . . . . . . . . . 14.2-12 14.2.6 Test Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-12 14.2.7 Test Programs Conformance with Regulatory Guides . . . . . . . . . . . . . 14.2-13 14.2.8 Utilization of Reactor Operating and Testing Experience in Test Program Development . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-14 14.2.9 Trial Use of Plant Operating Procedures, Emergency Procedures, and Surveillance Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-14 14.2.10 Initial Fuel Loading, and Initial Criticality . . . . . . . . . . . . . . . . . . . . . . 14.2-14 14.2.11 Test Program Schedule and Sequence. . . . . . . . . . . . . . . . . . . . . . . 14.2-18 14.2.12 Individual Test Descriptions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-19 14.3 Inspections, Tests, Analyses, and Acceptance Criteria . . . . . . . . . . . . . . . . 14.3-1 14.3.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-1 14.3.2 Top-Level Design Features and Inspections, Tests, Analyses, and Acceptance Criteria First Principles . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-1 14.3.3 Inspections, Tests, Analyses, and Acceptance Criteria Information . . . 14.3-6 14.3.4 Treatment of Module-Specific and Shared Structures, Systems, and Components in Inspections, Tests, Analyses, and Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.3-9 NuScale US460 SDAA i Revision 0

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 14.2-1: Test # 01 Pool Cooling and Cleanup System . . . . . . . . . . . . . . . . . . . . 14.2-20 Table 14.2-2: Test # 02 Ultimate Heat Sink . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-22 Table 14.2-3: Test # 03 Pool Leakage Detection System . . . . . . . . . . . . . . . . . . . . . . 14.2-23 Table 14.2-4: Test # 04 Reactor Component Cooling Water System . . . . . . . . . . . . . 14.2-24 Table 14.2-5: Test # 05 Chilled Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-26 Table 14.2-6: Test # 06 Auxiliary Boiler System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-28 Table 14.2-7: Test # 07 Air Cooled Condenser System. . . . . . . . . . . . . . . . . . . . . . . . 14.2-30 Table 14.2-8: Test # 08 Site Cooling Water System . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-32 Table 14.2-9: Test # 09 Potable Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-34 Table 14.2-10: Test # 10 Utility Water System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-35 Table 14.2-11: Test # 11 Demineralized Water System . . . . . . . . . . . . . . . . . . . . . . . . 14.2-37 Table 14.2-12: Test # 12 Nitrogen Distribution System . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-39 Table 14.2-13: Test # 13 Service Air System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-40 Table 14.2-14: Test # 14 Instrument Air System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-41 Table 14.2-15: Test # 15 Control Room Habitability System . . . . . . . . . . . . . . . . . . . . . 14.2-42 Table 14.2-16: Test # 16 Normal Control Room HVAC System . . . . . . . . . . . . . . . . . . 14.2-45 Table 14.2-17: Test # 17 Reactor Building HVAC System. . . . . . . . . . . . . . . . . . . . . . . 14.2-48 Table 14.2-18: Test # 18 Radioactive Waste Building HVAC System . . . . . . . . . . . . . . 14.2-51 Table 14.2-19: Test # 19 Turbine Building HVAC System . . . . . . . . . . . . . . . . . . . . . . . 14.2-53 Table 14.2-20: Test # 20 Radioactive Waste Drain System . . . . . . . . . . . . . . . . . . . . . 14.2-54 Table 14.2-21: Test # 21 Balance-of-Plant Drain System . . . . . . . . . . . . . . . . . . . . . . . 14.2-56 Table 14.2-22: Test # 22 Fire Protection System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-59 Table 14.2-23: Test # 23 Fire Detection System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-61 Table 14.2-24: Test # 24 Main Steam System. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-62 Table 14.2-25: Test # 25 Condensate and Feedwater System. . . . . . . . . . . . . . . . . . . . 14.2-63 Table 14.2-26: Test # 26 Feedwater Treatment System . . . . . . . . . . . . . . . . . . . . . . . . 14.2-65 Table 14.2-27: Test # 27 Condensate Polisher Resin Regeneration System . . . . . . . . . 14.2-66 Table 14.2-28: Test # 28 Feedwater Heater Vents and Drains System. . . . . . . . . . . . . 14.2-68 Table 14.2-29: Test # 29 Turbine Generator System. . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-70 Table 14.2-30: Test # 30 Liquid Radioactive Waste System . . . . . . . . . . . . . . . . . . . . . 14.2-73 Table 14.2-31: Test # 31 Gaseous Radioactive Waste System. . . . . . . . . . . . . . . . . . . 14.2-77 Table 14.2-32: Test # 32 Solid Radioactive Waste System . . . . . . . . . . . . . . . . . . . . . . 14.2-79 NuScale US460 SDAA ii Revision 0

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 14.2-33: Test # 33 Chemical and Volume Control System . . . . . . . . . . . . . . . . . 14.2-82 Table 14.2-34: Test # 34 Boron Addition System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-86 Table 14.2-35: Test # 35 Module Heatup System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-88 Table 14.2-36: Test # 36 Containment Evacuation System. . . . . . . . . . . . . . . . . . . . . . 14.2-89 Table 14.2-37: Test # 37 Containment Flooding and Drain . . . . . . . . . . . . . . . . . . . . . . 14.2-92 Table 14.2-38: Test # 38 Containment System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-94 Table 14.2-39: Test # 39 Reactor Coolant System . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-96 Table 14.2-40: Test # 40 Emergency Core Cooling System . . . . . . . . . . . . . . . . . . . . . 14.2-97 Table 14.2-41: Test # 41 Decay Heat Removal System . . . . . . . . . . . . . . . . . . . . . . . . 14.2-99 Table 14.2-42: Test # 42 In-Core Instrumentation System . . . . . . . . . . . . . . . . . . . . . 14.2-100 Table 14.2-43: Test # 43 Module Assembly Equipment . . . . . . . . . . . . . . . . . . . . . . . 14.2-101 Table 14.2-44: Test # 44 Fuel Handling Equipment. . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-102 Table 14.2-45: Test # 45 Reactor Building Cranes . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-104 Table 14.2-46: Test # 46 Process Sampling System . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-108 Table 14.2-47: Test # 47 High Voltage AC Electrical Distribution System. . . . . . . . . . 14.2-111 Table 14.2-48: Test # 48 Medium Voltage AC Electrical Distribution System . . . . . . . 14.2-112 Table 14.2-49: Test # 49 Low Voltage AC Electrical Distribution System . . . . . . . . . . 14.2-113 Table 14.2-50: Test # 50 Augmented DC Power System . . . . . . . . . . . . . . . . . . . . . . 14.2-114 Table 14.2-51: Test # 51 Normal DC Power System . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-117 Table 14.2-52: Test # 52 Backup Power Supply System. . . . . . . . . . . . . . . . . . . . . . . 14.2-120 Table 14.2-53: Test # 53 Plant Lighting System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-121 Table 14.2-54: Test # 54 Module Control System . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-122 Table 14.2-55: Test # 55 Plant Control System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-123 Table 14.2-56: Test # 56 Module Protection System . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-124 Table 14.2-57: Test # 57 Plant Protection System. . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-130 Table 14.2-58: Test # 58 Neutron Monitoring System . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-131 Table 14.2-59: Test # 59 Safety Display and Indication System . . . . . . . . . . . . . . . . . 14.2-132 Table 14.2-60: Test # 60 Fixed-Area Radiation Monitoring System . . . . . . . . . . . . . . 14.2-135 Table 14.2-61: Test # 61 Communication System . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-136 Table 14.2-62: Test # 62 Seismic Monitoring System . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-138 Table 14.2-63: Test # 63 Hot Functional Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-139 Table 14.2-64: Test # 64 Module Assembly Equipment Bolting . . . . . . . . . . . . . . . . . 14.2-142 NuScale US460 SDAA iii Revision 0

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 14.2-65: Test # 65 Steam Generator Flow-Induced Vibration . . . . . . . . . . . . . . 14.2-143 Table 14.2-66: Test # 66 Security Access Control. . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-144 Table 14.2-67: Test # 67 Security Detection and Alarm . . . . . . . . . . . . . . . . . . . . . . . 14.2-145 Table 14.2-68: Test # 68 Initial Fuel Loading and Precritical . . . . . . . . . . . . . . . . . . . . 14.2-146 Table 14.2-69: Test # 69 Initial Fuel Load . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-147 Table 14.2-70: Test # 70 Reactor Coolant System Flow Measurement . . . . . . . . . . . 14.2-148 Table 14.2-71: Test # 71 NuScale Power Module Temperatures . . . . . . . . . . . . . . . . 14.2-149 Table 14.2-72: Test # 72 Primary and Secondary System Chemistry . . . . . . . . . . . . . 14.2-150 Table 14.2-73: Test # 73 Control Rod Drive System - Manual Operation, Rod Speed, and Rod Position Indication. . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-151 Table 14.2-74: Test # 74 Control Rod Assembly Full-Height Drop Time . . . . . . . . . . . 14.2-152 Table 14.2-75: Test # 75 Control Rod Assembly Ambient Temperature Full-Height Drop Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-153 Table 14.2-76: Test # 76 Pressurizer Spray Bypass Flow . . . . . . . . . . . . . . . . . . . . . . 14.2-154 Table 14.2-77: Test # 77 Initial Criticality . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-155 Table 14.2-78: Test # 78 Post-Critical Reactivity Computer Checkout . . . . . . . . . . . . 14.2-156 Table 14.2-79: Test # 79 Low-Power Test Sequence . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-157 Table 14.2-80: Test # 80 Determination of Zero-Power Physics Testing Range . . . . . 14.2-158 Table 14.2-81: Test # 81 All Rods Out Boron Endpoint Determination . . . . . . . . . . . . 14.2-159 Table 14.2-82: Test # 82 Isothermal Temperature Coefficient Measurement . . . . . . . 14.2-160 Table 14.2-83: Test # 83 Bank Worth Measurement . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-161 Table 14.2-84: Test # 84 Power-Ascension . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-162 Table 14.2-85: Test # 85 Core Power Distribution Map . . . . . . . . . . . . . . . . . . . . . . . . 14.2-163 Table 14.2-86: Test # 86 Neutron Monitoring System Power Range Flux Calibration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-164 Table 14.2-87: Test # 87 Reactor Coolant System Temperature Instrument Calibration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-165 Table 14.2-88: Test # 88 Reactor Coolant System Flow Calibration . . . . . . . . . . . . . . 14.2-166 Table 14.2-89: Test # 89 Radiation Shield Survey. . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-167 Table 14.2-90: Test # 90 Reactor Building Ventilation System Capability . . . . . . . . . . 14.2-168 Table 14.2-91: Test # 91 Thermal Expansion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-169 Table 14.2-92: Test # 92 Control Rod Assembly Misalignment . . . . . . . . . . . . . . . . . . 14.2-170 Table 14.2-93: Test # 93 Steam Generator Level Control . . . . . . . . . . . . . . . . . . . . . . 14.2-171 NuScale US460 SDAA iv Revision 0

NuScale Final Safety Analysis Report List of Tables LIST OF TABLES Table 14.2-94: Test # 94 Ramp Change in Load Demand . . . . . . . . . . . . . . . . . . . . . 14.2-172 Table 14.2-95: Test # 95 Step Change in Load Demand. . . . . . . . . . . . . . . . . . . . . . . 14.2-174 Table 14.2-96: Test # 96 Loss of Feedwater Heater . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-175 Table 14.2-97: Test # 97 100 Percent Load Rejection. . . . . . . . . . . . . . . . . . . . . . . . . 14.2-176 Table 14.2-98: Test # 98 Reactor Trip from 100 Percent Power . . . . . . . . . . . . . . . . . 14.2-177 Table 14.2-99: Test # 99 Island Mode Test for the First NuScale Power Module . . . . 14.2-178 Table 14.2-100: Test # 100 Island Mode Test for Multiple NuScale Power Modules . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-179 Table 14.2-101: Test # 101 Remote Shutdown Controls and Monitoring . . . . . . . . . . . 14.2-180 Table 14.2-102: Test # 102 NuScale Power Module Vibration . . . . . . . . . . . . . . . . . . . 14.2-181 Table 14.2-103: List of Test Abstracts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14.2-182 Table 14.2-104: Initial Test Program Testing of New Design Features . . . . . . . . . . . . . . 14.2-185 NuScale US460 SDAA v Revision 0

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 US460 standard design. 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. The methodology associated with developing Inspections, Tests, Analyses, and Acceptance Criteria is described in Section 14.3.

The initial test program tests structures, systems, 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.

NuScale US460 SDAA 14.0-1 Revision 0

Specific Information to be Addressed for the Initial Plant Test NuScale Final Safety Analysis Report 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 tests and evaluate the results of tests as described in Section 14.2. The Initial Test Program is used to satisfy 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

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

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

10 CFR 52.79 as it relates 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 NuScale US460 SDAA 14.1-1 Revision 0

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 conducted by the Startup organization. Preoperational testing is conducted for each NuScale Power Module (NPM) following completion of construction testing but before fuel load. Completion of preoperational testing for each NPM is necessary to ensure the NPM is ready for fuel loading and startup testing.

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 practicable the NPM operates in accordance with 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 is completed in accordance with the design.

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

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

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

verify to the extent practicable by trial use the facility operating procedures, that 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.

NuScale US460 SDAA 14.2-1 Revision 0

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

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 analyses 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 Criterion XI of 10 CFR 50 Appendix B. Implementation of the ITP ensures testing required to demonstrate an 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.

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

The methodology associated with the development of the ITAAC necessary to demonstrate the facility is constructed and is operated in conformity with the Final Safety Analysis Report (FSAR) and the applicable Nuclear Regulatory Commission (NRC) regulations is presented in Section 14.3.

The designs compliance with the proposed technical resolution of unresolved safety issues and medium- and high-priority generic safety issues 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 1.9.3.

14.2.1.1 Construction Organization Testing The objective of construction testing is to verify, on a system basis, the system is constructed and installed in accordance with design requirements. Construction tests include but are not limited to

flushing.

cleaning.

hydrostatic pressure tests.

NuScale US460 SDAA 14.2-2 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program

wiring continuity and separation checks.

The testing and installation of digital instrumentation and controls (I&C) systems are described in Section 7.2.1, and includes factory acceptance testing and site acceptance testing (SAT) completed as part of construction and installation tests performed before, and as a prerequisite of, preoperational tests. Factory acceptance tests are performed during the digital I&C system testing phase described in Section 7.2.1. Site installation and checkout activities are performed as part of the integrated SAT during the system installation phase as described in Section 7.2.1. Software integration and testing is governed by the Digital I&C Software Master Test Plan described in Section 7.2.1.

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

demonstrate SSC 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 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 design, manufacturing, and installation defects can be detected and corrected.

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

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

demonstrate equipment performance.

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

NuScale US460 SDAA 14.2-3 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program

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

provide the plant operating staff with the opportunity to obtain practicable 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.

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 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. The objectives of the initial fuel loading and pre-critical tests are the following:

conduct initial fuel loading cautiously to preclude inadvertent criticality

establish and follow specific safety measures, such as ensuring the applicable TS requirements and other prerequisites are satisfied.

continuous monitoring of the neutron flux throughout core loading so changes in the multiplication factor are observed.

verifying fuel and control components are properly installed.

establish the required shutdown margin 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 NuScale US460 SDAA 14.2-4 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program 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 as follows. 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 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 is registered on the startup channels before startup begins, and the signal to noise ratio is 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.

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 TS 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.

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

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 system (MSS).

NuScale US460 SDAA 14.2-5 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program 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 the plant operates in accordance with its design bases during normal steady-state conditions and, to the extent practicable, 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 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 (FWHs) or pumps.

14.2.2 Organization and Staffing COL Item 14.2-1: An applicant that references the NuScale Power Plant US460 standard design 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. After the test procedures are 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 FSAR, 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 NuScale US460 SDAA 14.2-6 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program

application, to the extent practicable, 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 practicable to those the equipment experiences in both normal and accident situations.

methods to direct and control test performance.

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 are 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 (e.g., 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, SSC 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 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 NuScale US460 SDAA 14.2-7 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program allow for testing of SSC in a manner commensurate with the safety significance of the SSC within its scope. These provisions provide a systematic approach to the defense-in-depth concept. This concept dictates the plant be designed, constructed, and tested to

provide for safe normal operation,

ensure, in the event of errors, malfunctions, and off-normal conditions, the reactor protection systems and other design features mitigate the event or limit consequences to defined and acceptable levels, and

ensure 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 the SSC being tested 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 (i.e., procedures and records) associated with testing to be prepared commensurate with the safety significance of the SSC being tested.

During the SSC classification process, the subject matter expert identifies functions of the system. Each of these functions is compared to safety functional requirements and regulatory functional requirements to establish a functional hierarchy. This hierarchy establishes a relationship among the systems and ties it to a set of plant functions as described in Section 17.4 to identify a classification for the functions. The functions are 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 is the basis for the graded approach in the ITP.

The hierarchy in preoperational testing is

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

testing of active, nonsafety-related functions that require ITAAC verification (B1 and B2).

testing of active nonsafety-related functions that do not require ITAAC verification (B1 and B2).

The preoperational test abstracts contained in Table 14.2-1 through Table 14.2-62 define 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 A1 and B1 system functions. Active, safety-related A1 functions are tested by the safety-related module protection system (MPS) logic testing found in Table 14.2-56. The remaining safety-related functions categorized as A2 are also tested by the MPS test abstract. The graded approach provides for testing of A2 functions to the same rigor as A1 functions.

NuScale US460 SDAA 14.2-8 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program As indicated by Table 14.2-56, 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 (PZR) 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 safety-related containment isolation valve (CIV) response time and MPS safety-related sensor response time.

Section 14.3 provides guidance regarding the development of ITAAC. The successful completion of ITAAC constitutes the basis for the NRC determination to allow operation of a facility licensed under 10 CFR 52. The ITAAC are verified by an inspection, test, or analysis, or a combination thereof. Some ITAAC are verified by successful completion of preoperational testing.

Each ITAAC is identified by its unique ITAAC number, for example, ITAAC 03.01.02. If an ITAAC is verified by the successful completion of a preoperational test, the acceptance criteria of the associated test in Section 14.2 contain a bracketed reference to the verified ITAAC. An example annotation is

[ITAAC 03.01.02] in Table 14.2-15 where 03.01.02 is the number of the verified ITAAC.

Credit is taken for the logic testing performed for the nonsafety-related module control system (MCS) described in Section 7.2.1, and the nonsafety-related plant control system (PCS) described in Section 7.2.1. 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 systems. This graded approach does not affect system-level tests that 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

equipment response to automatic signals to protect plant equipment NuScale US460 SDAA 14.2-9 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program

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 can be monitored in the MCR (test not required if the instrument calibration verified the MCR display) 14.2.3.3 Testing of First-of-a-Kind Design Features First-of-a-kind (FOAK) tests are new, unique, or special tests to verify design features reviewed for the first time by the NRC. The NuScale Power Plant contains design features that are new and unique and not 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 Comprehensive Vibration Assessment Program (CVAP) is an FOAK program. The program is implemented consistent with the requirements of the NuScale Comprehensive Vibration Assessment Program Technical Report," TR-121353-P, and the NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report, TR-121354-P. 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 14.2-65: Steam Generator Flow-Induced Vibration Test #65

Table 14.2-102: NuScale Power Module Vibration Test #102 The test results for the CVAP testing of the first NPM inform the required CVAP testing on subsequent NPMs as described in TR-121353-P. Other ITP testing of FOAK design features is performed for each NPM, except as described below.

Table 14.2-40: Emergency Core Cooling System Test #40 includes a one-time in-situ system performance test of the emergency core cooling system (ECCS).

The test demonstrates valve and containment response to manual emergency safety feature actuation of the ECCS at hot functional test pressure and temperature.

Section 5.4.3 contains a description of the decay heat removal system (DHRS) one-time in-situ RCS heat removal test. The test is performed per test abstract Table 14.2-41: Decay Heat Removal System Test # 41.

Table 14.2-104 provides a summary of the ITP testing (i.e., preoperational and startup testing) for new design features. Each test is performed for all NPMs.

Section 1.5.1 contains a description of testing programs that have been completed or are currently in progress for design features for which applicable data or operational experience did not previously exist. The section describes NuScale US460 SDAA 14.2-10 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program tests specific to fuel design, steam generator (SG), and control rod assemblies (CRAs).

14.2.3.4 Generic Component Testing Component testing is generally executed after a system transfers from the construction organization to the startup organization. Generic component testing executes standardized tests for a family of related component types, independent of the components system assignment. Each generic component test procedure is completed and approved before the component is required as a prerequisite to a preoperational test performance. The completion of generic component testing is listed as a prerequisite in each preoperational test procedure as applicable.

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

COL Item 14.2-2: An applicant that references the NuScale Power Plant US460 standard design will develop the Startup Administration Manual that will contain the administrative procedures and requirements that control the activities associated with the Initial Test Program. The applicant will provide a milestone for completing the Startup Administrative Manual and making it available for Nuclear Regulatory Commission inspection.

Administrative controls are established to ensure designated construction-related inspections and tests are completed before initiating preoperational testing. In addition, controls are established to ensure completion of preoperational testing before 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 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 are conducted before performing the modification or repair.

Controls are established to ensure retesting 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.

NuScale US460 SDAA 14.2-11 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program 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. These procedures include 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. 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

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 Specifications,"

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, 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 NuScale US460 SDAA 14.2-12 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program

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, Initial Test Programs for Water-Cooled Nuclear Power Plants, 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 relative to testing of SSC:

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

RG 1.29 - Seismic Design Classification for Nuclear Power Plants

RG 1.41 - Preoperational Testing of Redundant On-Site Electric Power Systems to Verify Proper Load Group Assignments

RG 1.45 - Guidance on Monitoring and Responding to Reactor Coolant System Leakage

RG 1.68.1 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

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

RG 1.68.3 - Preoperational Testing of Instrument and Control Air Systems

RG 1.69 - Concrete Radiation Shields and Generic Shield Testing for Nuclear Power Plants

RG 1.79 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors

RG 1.118 - Periodic Testing of Electric Power and Protection Systems

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

RG 8.38 - Control of Access to High and Very High Radiation Areas of Nuclear Power Plants Refer to Section 1.9 for information related to the conformance with each of these regulatory guides.

NuScale US460 SDAA 14.2-13 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program 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., 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 into 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 practicable, developed, trial tested, and corrected during the ITP before fuel load to establish their adequacy. Trial testing of procedures is accomplished by training plant operators to these procedures to the extent practicable during the ITP. Following completion of trial testing these procedures are used as part of the ITP.

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 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 before 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.

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 is performed by first implementing the prerequisite and precautionary measures contained in test procedures and identified below:

TS compliance is met

successful completion of all ITAAC NuScale US460 SDAA 14.2-14 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program

identification of actions to be taken in the event of unanticipated errors or malfunctions

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 performed because of 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

ensuring applicable TS are met,

performing continuous monitoring of the neutron flux throughout core loading so changes in the multiplication factor are observed,

establishing requirements for periodic data taking, and

independently verifying fuel and control components are 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 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 before the initial fuel loading. In addition, before initial fuel loading the required shutdown margin 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 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.

NuScale US460 SDAA 14.2-15 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program The following conditions exist before 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 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 trips are set at the lowest value.

The Radiation Monitoring Program as it pertains to operation of radiation barriers, airborne radiation monitors, and air sampling is implemented.

Baseline surveys are performed before withdrawing control rods for the approach to critical.

14.2.10.4 Low-Power Testing Following initial criticality, low-power tests (at less than 5 percent power) are conducted to confirm the design, validate the analytical models to the extent practicable, and verify the correctness or conservatism of assumptions used in the safety analyses for the facility. Low-power tests are also used to confirm the functionality of plant systems and design features not completely tested during the preoperational test phase because of the lack of an adequate heat source from the RCS and MSS.

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, measurements 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.

Section 14.2.12 contains a list of low-power tests.

NuScale US460 SDAA 14.2-16 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program COL Item 14.2-3: An applicant that references the NuScale Power Plant US460 standard design will identify the specific operator training to be conducted during low-power testing related to the resolution of Three Mile Island 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 the plant operates in accordance with its design bases during normal steady state conditions and, to the extent practicable, during and following anticipated transients. Testing is also intended 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 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 trip, turbine trip, and loss of FWHs or pumps. The testing performed is sufficiently comprehensive to establish the facility can operate in the operating modes for which it is designed. Testing is conducted in operating modes or plant configurations analyzed, and are bounded by the range of assumptions used in analyzing postulated accidents as described in the FSAR.

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, measurements taken and test conditions as well as actions taken in the event of unanticipated errors or malfunctions. These procedures provide direction for restoration to normal following the test.

Section 14.2.12 contains 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.

NuScale US460 SDAA 14.2-17 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program 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 SSC required to prevent or mitigate the consequences of postulated accidents are tested before fuel loading.

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

Test procedures are essentially identical for each NPM. Structures, systems, and components identification numbering is specific to each NPM.

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

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

Vibration testing is performed in accordance with the CVAP as described in the "NuScale Comprehensive Vibration Assessment Program Technical Report,"

TR-121353-P. Test results are verified following power-ascension testing.

Section 3.9.2 contains information pertaining to the CVAP.

The sequential schedule for individual startup tests establishes that, insofar as practicable, test requirements are completed before exceeding 25 percent power for the plant SSC relied upon to prevent, limit, or mitigate the consequences of postulated accidents. The schedule establishes, insofar as practicable, the sequencing of testing is accomplished as early in the test program as feasible and the safety of the plant is not dependent on the performance of untested systems, components, or features. Startup test data are reviewed and approved before moving onto the next power plateau. Startup testing is discussed in Section 14.2.1.

The plant is comprised of up to six NPMs. A schedule is developed for startup of each NPM. Preoperational and startup testing schedule considerations include

preoperational test schedule duration is the longest for the first NPM because the first NPM requires testing of systems common to other NPMs.

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

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

plant safety must not be dependent on the performance of untested SSC during the startup test program.

NuScale US460 SDAA 14.2-18 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Section 1.1.4 contains information pertaining to phased construction and testing activities due to addition of individual NPMs.

COL Item 14.2-4: An applicant that references the NuScale Power Plant US460 standard design 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 14.2-102.

Table 14.2-103 provides a list of the test abstracts. Each abstract identifies each test by title, and 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).

NuScale US460 SDAA 14.2-19 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-1: Test # 01 Pool Cooling and Cleanup System Preoperational test is required to be performed once.

The pool cooling and cleanup system (PCWS) comprises three subsystems, pool cooling, pool cleanup, and pool surge control described in Section 9.1.3 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The PCWS supports pool cleanup nonsafety-related 01.02.01 by providing fuel pool water for purification of the ultimate heat sink (UHS).

2. The PCWS supports pool cleanup nonsafety-related 01.02.01 by providing reactor pool water for purification of the UHS.

3. The PCWS supports pool cleanup nonsafety-related 01.02.02 by providing water from the dry Component level tests dock for UHS inventory control.

4. The PCWS supports the UHS by nonsafety-related 01.02.02 providing surge control for UHS operations.

5. The PCWS supports the UHS by nonsafety-related 01.02.02 providing a reactor inspection dry dock makeup and drain capability.

6. The PCWS supports pool surge nonsafety-related Component level tests control, reactor pool cooling and fuel pool cooling by providing a flowpath to cross-connect pool surge control, reactor pool cooling, and fuel pool cooling subsystems.

01.00.XX Prerequisites

01. Verify instrument calibration is complete, with approved records and within calibration due dates, for instruments required to perform this test.

02. Verify a pump curve test is complete and approved for the PCWS pumps.

03. Verify a UHS leakage test is complete.

04. For system level test 01.02.02, the dry dock gate can be in the open or closed position.

01.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each PCWS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each PCWS air-operated 1) Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. Isolate and vent air to the observation indicate each valve loss of air. valve. fails to its safe position.

03. Verify each PCWS air-operated 1) Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. Isolate electrical power to observation indicate each valve loss of electrical power to its each air-operated valve. fails to its safe position.

solenoid.

04. Verify each PCWS pump can be 1) Align the PCWS to allow for pump 1) MCR display and local, visual started and stopped remotely. operation. Start and stop each observation indicate each pump pump from the MCR. starts and stops.

05. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from each PCWS grab flow through the grab sampling obtained from each device.

sample device. device.

NuScale US460 SDAA 14.2-20 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-1: Test # 01 Pool Cooling and Cleanup System (Continued)

06. Verify each PCWS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each PCWS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

01.02.XX System Level Tests 01.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the pool cooling subsystem 1) 1) provides design flow rate to the a. Place fuel pool cooling in service to a. The MCR indication for PCWS UHS when aligned for pool flow through a pool cleanup filter pump flow satisfies the design flow cleanup. and demineralizer and return flow rate specified in Table 9.1.3-1.

2. Verify the pool cooling subsystem to the fuel pool. b. The MCR indication for PCWS provides design flow rate to the AND demineralizer flow satisfies the UHS following a pool cleanup b. Place reactor pool cooling in design flow rate specified in isolation. service to flow through a different Table 9.1.3-2.

pool cleanup filter and 2) demineralizer and return flow to the a. PCWS flow to the pool cleanup reactor pool. filters and demineralizers stop.

2) Simulate a high water temperature b. The PCWS flow is bypassed to the upstream of one of the pool fuel pool.

cleanup filters. c. The PCWS flow is bypassed to the reactor pool.

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

01.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify PCWS automatic control for Align the PCWS for fill and drain of the 1) Verify inventory addition to the dry dock fill and drain. dry dock. PCWS surge and control tank.

Fill the dry dock to a level that allows 2) operation of the dry dock evacuation a. Pump is stopped and return line to pump. pool surge control tank isolation

1) Start a PCWS dry dock evacuation valve is closed.

pump. b. Pump is stopped and return line to

2) Simulate the following PCWS PCWS surge control and surge conditions: tank isolation valve is closed.

a. Dry dock low level 3) Verify inventory addition to the dry

b. PCWS surge control and storage dock.

tank high level 4) PCWS surge control and surge

3) Open PCWS surge and control tank main discharge line isolation tank main discharge line isolation valve is closed.

valve.

4) Simulate a high dry dock level.

NuScale US460 SDAA 14.2-21 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-2: Test # 02 Ultimate Heat Sink 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). Safety Display and Indication Test 59.02.02 discusses testing of PAM Type D displays.

System Function System Function Categorization Function Verified by Test #

1. The UHS supports the DHRS by safety-related 98.03.01 accepting the heat from the DHRS heat exchanger.

02.00.XX Prerequisites:

N/A 02.01.XX Component Level Tests None NuScale US460 SDAA 14.2-22 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-3: Test # 03 Pool Leakage Detection System There are no preoperational tests for the pool leakage detection system (PLDS).

The PLDS is described in Section 9.1.3. Leakage from the UHS drains via gravity to the radioactive waste drain system (RWDS). Test 20.02.02 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 03.00.XX Prerequisites:

N/A 03.01.XX Component Level Tests None NuScale US460 SDAA 14.2-23 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-4: Test # 04 Reactor Component Cooling Water System Preoperational test is required to be performed once each for the shared or common components. The module-specific portions of the test must be completed once for each NPM.

The reactor component cooling water system (RCCWS) is described in Section 9.2.2 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

1. The RCCWS supports the nonsafety-related 04.02.01 following systems by providing cooling water.

control rod drive system (CRDS)

CVCS

containment evacuation system (CES)

process sampling system (PSS) 04.00.XX Prerequisites

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

02. Verify an RCCWS flow balance is performed.

03. Verify a pump curve test is completed for the RCCWS pumps.

04.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each RCCWS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each RCCWS air-operated 1) Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. Isolate and vent air to the observation indicate each valve loss of air. valve. fails to its safe position.

03. Verify each RCCWS air-operated 1) Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. Isolate electrical power to observation indicate each valve loss of electrical power to its each air-operated valve. fails to its safe position.

solenoid.

04. Verify each RCCWS pump can be 1) Align the RCCWS to allow for 1) MCR display and local, visual started and stopped remotely. pump operation. Start and stop observation indicate each pump each pump from the MCR. starts and stops. Audible and visible water hammer are not observed when the pump starts.

05. Verify the RCCWS standby pump 1) Align the RCCWS to allow for 1) MCR display and local, visual automatically starts to protect pump operation. Place a pump in observation indicate the standby plant equipment. service. Initiate a simulated pump starts. Audible and visible RCCWS pump low header water hammer are not observed pressure signal. when the pump starts.

06. Verify RCCWS demineralized 1) Initiate simulated expansion tank 1) MCR display and local, visual makeup water level control valve high level signal. observation indicate the following:

automatically operates to maintain The demineralized makeup water RCCWS expansion tank level. level control valve is fully closed.

07. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from an RCCWS grab flow through the grab sampling obtained.

sample device. device.

08. Verify each RCCWS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each on an MCS or PCS display, or is display. RCCWS transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

NuScale US460 SDAA 14.2-24 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-4: Test # 04 Reactor Component Cooling Water System (Continued) 04.02.XX System Level Test 04.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify RCCWS cooling water flow Module 1 Test 1) The RCCWS cooling flow to each rates satisfy design flow. 1) Align RCCWS to provide flow to all heat exchanger under test meets module 1 heat exchangers listed the flow rate acceptance criteria below: contained in the RCCWS flow

control rod drive mechanism balance report.

(CRDM) cooling coils

CVCS non-regenerative heat exchanger

CES vacuum pump

CES condenser

PSS analyzer cooler

PSS temperature control unit

2) Repeat module 1 test for modules 2 through 6.

NuScale US460 SDAA 14.2-25 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-5: Test # 05 Chilled Water System 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 #

1. The CHWS supports the following nonsafety-related 05.02.01 systems by providing cooling 05.02.02 water:

Reactor Building HVAC system (RBVS)

normal control room HVAC system (CRVS)

Radioactive Waste Building (RWB) HVAC system (RWBVS)

liquid radioactive waste system (LRWS)

gaseous radioactive waste system (GRWS) 05.00.XX Prerequisites

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

02. Verify a CHWS flow balance is performed.

03. Verify a pump curve test is completed for the CHWS pumps.

04 Chiller performance is verified by either an Air Conditioning, Heating, and Refrigeration Institute (AHRI) certification or a chiller performance capacity test witnessed at the factory with all test documentation provided.

05.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each CHWS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify the speed of each CHWS 1) Align the CHWS to provide a flow 1) MCR display indicates the speed variable-speed pump can be path to operate a selected pump. of each pump obtains both manually controlled. Vary the CHWS pump speed from minimum and maximum pump minimum to maximum from the speeds. Audible and visible water MCR. hammer are not observed when the pump starts.

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

to protect plant equipment. Initiate a simulated signal for the 1) Operating pump stops following system conditions. 2) Operating chiller stops

1) Loss of chilled water flow.

2) Loss of site cooling water system (SCWS) cooling flow to the operating chiller.

04. Verify each CHWS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each CHWS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

NuScale US460 SDAA 14.2-26 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-5: Test # 05 Chilled Water System (Continued) 05.02.XX System Level Tests 05.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify CHWS cooling water flow 1) Align the CHWS to provide flow to 1) The CHWS cooling flow to each rates satisfy design. all heat exchangers cooled by the heat exchanger under test meets CHWS chiller: the minimum flow rate acceptance

RBVS air handling units (AHUs) criteria contained in the CHWS

RBVS fan coil units flow balance report.

CRVS AHUs

CRVS fan coil units

RWBVS AHU

RWBVS fan coil units

LRWS degasifier condenser

GRWS gas coolers

2) Open all CHWS flow control valves.

05.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify CHWS cooling water flow 1) Align the CHWS to provide flow to 1) The CRVS standby CHWS cooling rates satisfy design flow. the CRVS AHUs and the CRVS fan flow to each heat exchanger coil units cooled by the CHWS meets the minimum flow rate standby chiller. acceptance criteria contained in

2) Open all CHWS flow control the CHWS flow balance report.

valves.

NuScale US460 SDAA 14.2-27 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-6: Test # 06 Auxiliary Boiler System Preoperational test is required to be performed once unless specified otherwise in the test method.

The auxiliary boiler system (ABS) is described in Section 10.4.7 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:

1. The auxiliary boiler supports the nonsafety-related Table 14.2-27 Component Level condensate polisher resin Tests regeneration system (CPS) by supplying chemicals for neutralization.

2. The ABS supports the turbine nonsafety-related 07.02.01 generator by supplying gland seal steam.

3. The ABS supports the air cooled nonsafety-related 07.02.01 condenser system (ACCS) by supplying steam to the condenser deaerator during startup.

06.00.XX Prerequisites

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

06.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each auxiliary boiler 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each auxiliary boiler 1) Place each valve in its non-safe 1) MCR display and local, visual air-operated valve fails to its safe position. Isolate and vent air to the observation indicate each valve position on loss of air. valve. fails to its safe position.

03. Verify each auxiliary boiler 1) Place each valve in its non-safe 1) MCR display and local, visual air-operated valve fails to its safe position. Isolate electrical power to observation indicate each valve position on loss of electrical power each air-operated valve. fails to its safe position.

to its solenoid.

04. Verify each auxiliary boiler Align the ABS to allow for pump 1) MCR display and local, visual feedwater (FW) pump can be operation. observation indicate each pump started and stopped remotely. 1) Start and stop each pump from the starts and stops. Audible and MCR. visible water hammer are not observed when the pump starts.

NuScale US460 SDAA 14.2-28 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-6: Test # 06 Auxiliary Boiler System (Continued)

05. Verify the ABS automatically 1) For the first NPM being tested, 1) MCR display verifies the following responds to mitigate a release of initiate a real or simulated signal for valves are closed:

radioactivity. each of the following radiation a. Auxiliary boiler superheater skid trasmitters: outlet isolation.

auxiliary boiler skid vent b. Auxiliary boiler skid to superheater

auxiliary boiler to superheater ski skid isolation.

BPDS outlet c. Main steam to auxiliary boiler

auxiliary boiler to superheater header isolation.

skid vent 2) MCR display verifies the Main

auxiliary steam to BPDS steam to auxiliary boiler header

TGB auxiliary steam header isolation for the module being tested is

2) For each additional module, initiate closed.

a real or simulated high radiation [ITAAC 03.09.08] (items 1 and 2) signal to the specific valve being [ITAAC 02.07.03] (items 1 and 2) tested for each of the following radiation transmitters:

auxiliary boiler skid vent

auxiliary boiler to superheater skid BPDS outlet

auxiliary boiler to superheater skid vent

auxiliary steam to BPDS

TGB auxiliary steam header

06. Verify each ABS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each ABS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

06.02.XX System Level Test Test Objective Test Method Acceptance Criteria None NuScale US460 SDAA 14.2-29 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-7: Test # 07 Air Cooled Condenser System Preoperational test is required to be performed once each for the shared or common components. The module-specific portions of the test must be completed once for each NPM.

The Air Cooled Condenser System (ACCS) is described in Section 10.4.1 and the functions verified by this test and power ascension testing are:

System Function System Function Categorization Function Verified by Test #

1. The demineralized water system nonsafety-related Component-Level Test (DWS) and condensate storage 94.03.01 tank support the ACCS by providing makeup water to maintain water level in the condensate collection tank (CCT).

2. The ACCS supports the FWS by nonsafety-related 07.02.01 removing heat from the steam turbine exhaust and turbine bypass.

3. The ABS supports the turbine nonsafety-related 07.02.01 generator by supplying gland seal steam.

4. The ABS supports the ACCS by nonsafety-related 07.02.01 supplying steam to the condenser deaerator during startup.

07.00.XX Prerequisites

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

07.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each ACCS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each ACCS air-operated 1) Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. Isolate and vent air to the observation indicate each valve loss of air. valve. fails to its safe position.

03. Verify each ACCS air-operated 1) Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. Isolate electrical power to observation indicate each valve loss of electrical power to its each air-operated valve. fails to its safe position.

solenoid.

04. Verify each ACCS fan can be 1) Align the ACCS to allow for fan 1) MCR display and local, visual started and stopped remotely. operation. observation indicate each ACCS

2) Start and stop each ACCS fan from fan starts and stops.

the MCR.

05. Verify the fan speed of each 1) Vary the speed of each fan from 1) MCR display indicates the speed variable speed fan can be the MCR and local control panel (if of each fan varies from minimum to controlled manually. design has local fan control). maximum fan speed.

06. Verify each ACCS pump can be 1) Align the ACCS to allow for pump 1) MCR display and local, visual started and stopped remotely. operation. observation indicate each pump

2) Start and stop each pump from the starts and stops. Audible and MCR. visible water hammer are not observed when the pump starts.

2) ACCS pump cavitation is not observed.

NuScale US460 SDAA 14.2-30 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-7: Test # 07 Air Cooled Condenser System (Continued)

07. Verify automatic operation of the 1) Initiate a CCT low level signal. MCR displays and local, visual ACCS level control valves to 2) Initiate a CCT high level signal. observation verifies the following:

maintain CCT level. 1) The makeup level control valve is open.

2) The reject level control valve is opened.

08. Verify a local integrated grab 1) Place the system in service to allow 1) A local grab sample is obtained.

sample can be obtained from each flow through the grab sampling ACCS grab sample device. device.

09. Verify each ACCS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each ACCS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

07.02.XX System Level Tests 07.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the condenser air removal Place the ABS in automatic control to 1) Maintain main condenser design system (CARS) can maintain main supply gland seal steam. vacuum pressure.

condenser vacuum pressure. Place the FWS in automatic control to 2) The ABS is capable of providing condense the gland seal steam in the steam to the condenser deaerator gland exhaust condenser. as indicated by steam flow.

Place the ACCS in automatic control to 3) The ABS is capable of supplying provide cooling to the main condenser. gland seal steam to the turbine

1) Place the ACCS in service to generator at design pressures.

establish vacuum in the main condenser.

2) Open the steam supply isolation valves to provide steam to the condenser deaerator (DEA).

NuScale US460 SDAA 14.2-31 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-8: Test # 08 Site Cooling Water System Preoperational test is required to be performed once each for the shared or common components. The module-specific portions of the test must be completed once for each NPM.

The SCWS is described in Section 9.2.7 and the functions verified by this test and power ascension testing are:

System Function System Function Categorization Function Verified by Test #

1. The SCWS supports the following nonsafety-related 08.02.01 systems by providing cooling water.

turbine generator system (TGS)

RCCWS

ACCS

PSS

CHWS

instrument and control air system (IAS)

PCWS

FWS 08.00.XX Prerequisites

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

02. Verify an SCWS flow balance is performed and the system flow balance records have been approved.

03. Verify a pump curve test is completed and approved for the SCWS pumps.

08.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each SCWS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each SCWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each SCWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify the SCWS automatically 1) Initiate a real or simulated high 1) MCR display verifies the SCWS responds to mitigate a release of radiation signal for the SCWS blowdown isolation valve is radioactivity blowdown line closed.

[ITAAC 03.09.11]

05. Verify each SCWS cooling tower Align the SCWS to allow for cooling 1) MCR display and local, visual fan can be started and stopped tower fan operation. observation indicate each cooling remotely. 1) Start and stop each cooling tower tower fan starts and stops.

fan from the MCR.

06. Verify each SCWS pump can be Align the SCWS to allow for pump 1) MCR display and local, visual started and stopped remotely. operation. observation indicate each pump

1) Start and stop each pump from the starts and stops. Audible and MCR. visible water hammer are not observed when the pump starts.

07. Verify the SCWS standby pump Align the SCWS to allow for pump 1) MCR display and local, visual automatically starts to protect operation. observation indicate the standby plant equipment. 1) Place a pump in service. Initiate a pump starts. Audible and visible simulated start signal. water hammer are not observed when the pump starts.

08. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from an SCWS grab flow through the grab sampling obtained.

sample device. device.

NuScale US460 SDAA 14.2-32 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-8: Test # 08 Site Cooling Water System (Continued)

09. Verify each SCWS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each SCWS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

08.02.XX System Level Test 08.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify SCWS cooling water flow The operation of two SCWS pumps 1) The SCWS cooling flow to each rates satisfy design flow. may be required to provide sufficient heat exchanger under test meets flow to meet acceptance criteria in the the minimum flow rate acceptance SCWS flow balance report. criteria contained in the SCWS

1) For the first NPM tested: flow balance report.

Align the SCWS to provide flow to all of the first module and common heat exchangers cooled by SCWS listed below:

Module Heat Exchangers TGS coolers PSS coolers Common Heat Exchangers CHWS chillers IAS coolers PSS chillers PCWS heat exchangers ACC liquid ring vacuum pump skid heat exchangers RCCWS heat exchangers

2) Subsequent NPM tests The scope of each subsequent test includes one or more additional modules. The scope also includes 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 continues until all heat exchangers cooled by SCWS have been tested in a single test.

NuScale US460 SDAA 14.2-33 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-9: Test # 09 Potable Water System 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 applicant.

COL Item 14.2-5: An applicant that references the NuScale Power Plant US460 standard design 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 applicant 09.00.XX Prerequisites Provided by applicant 09.01.XX Component Level Tests Provided by applicant 09.02.XX System Level Tests Provided by applicant NuScale US460 SDAA 14.2-34 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-10: Test # 10 Utility Water System Preoperational test is required to be performed once.

The utility water system (UWS) is described in Section 9.2.9 and the functions verified by this test and power ascension testing are:

System Function System Function Categorization Function Verified by Test #

1. The UWS supports the SCWS by nonsafety related 08.01.01 providing makeup water to 94.03.01 maintain water level in the SCWS cooling tower basins.

2. The UWS supports the following nonsafety-related component-level tests systems by providing makeup water:

DWS

fire protection system (FPS)

PWS

CHWS

RXB

Turbine Generator Building (TGB)

RWB

Annex Building

CRB 10.00.XX Prerequisites

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

02. Verify a pump curve test is completed for the UWS pumps.

10.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each UWS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each UWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each UWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify each UWS pump can be Align the UWS to allow for pump 1) MCR display and local, visual started and stopped remotely. operation. observation indicate each pump

1) Start and stop each pump from the starts and stops. Audible and MCR. visible water hammer are not observed when the pump starts.

05. Verify UWS flow capability by Align the UWS to allow for pump 1) MCR display and local, visual automatic start of a UWS pump operation. Place a pump in service. observation indicate the standby while in standby mode. 1) Initiate a simulated pump trip. pump starts. Audible and visible water hammer are not observed when the pump starts.

06. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is obtained.

obtained from each UWS grab flow through the grab sampling sample device. device.

NuScale US460 SDAA 14.2-35 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-10: Test # 10 Utility Water System (Continued)

07. Verify each UWS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each UWS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

10.02.XX System Level Tests None NuScale US460 SDAA 14.2-36 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-11: Test # 11 Demineralized Water System Preoperational test is required to be performed once.

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

System Function System Function Categorization Function Verified by Test #

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

CVCS

boron addition system (BAS)

LRWS

PCWS

RCCWS

PSS

FWS

ABS

CRVS

ACCS

CES

RBVS

RWBVS

condensate polishing system

Annex Building (ANB)

balance-of-plant drain system (BPDS)

Turbine Building HVAC system (TBVS)

annex building HVAC system

RXB

RWB

feedwater treatment system (FWTS) 11.00.XX Prerequisites

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

02. Verify a pump curve test is completed for the DWS pumps.

11.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each DWS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control) fully opens and fully closes.

02. Verify each DWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each DWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify the DWS automatically 1) Initiate a real or simulated high 1) MCR display verifies the following responds to mitigate a release of radiation signal for each or the valves are closed.

radioactivity following: a. North Reactor Building

a. North Reactor Building demineralized water distribution demineralized water distribution header isolation valve header b. South Reactor Building

b. South Reactor Building demineralized water distribution demineralized water distribution header isolation valve header [ITAAC 03.09.09]

NuScale US460 SDAA 14.2-37 Revision 0

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

05. Verify the DWS pump can be Align the DWS to allow for pump 1) MCR display and local, visual started and stopped remotely. operation. observation indicate each pump

1) Start and stop each pump from the starts and stops. Audible and MCR. visible water hammer are not observed when the pump starts.

06. Verify DWS flow capability by Align the DWS to allow for pump 1) MCR display and local, visual automatic start of a DWS pump operation. Place a pump in service. observation indicate the standby while in standby mode. 1) Initiate a simulated pump trip. pump starts. Audible and visible water hammer are not observed when the pump starts.

07. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is obtained.

obtained from each DWS grab flow through the grab sampling sample device. device.

08. Verify each DWS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each DWS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

11.02.XX System Level Tests None NuScale US460 SDAA 14.2-38 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-12: Test # 12 Nitrogen Distribution System 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 #

1. The NDS supports the following nonsafety-related component-level tests systems by providing nitrogen:

CVCS

CES

RXB

2. The NDS supports the LRWS by nonsafety-related component-level tests providing nitrogen for purging of 30.02.01 the LRWS.

3. The NDS supports the GRWS by nonsafety-related component-level tests providing nitrogen for purging of 31.02.01 the GRWS.

12.00.XX Prerequisites 01.Verify an instrument calibration is completed, with approved records and within calibration due dates, for instruments required to perform this test.

12.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each NDS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each NDS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each NDS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from a NDS grab sample flow through the grab sampling obtained.

device. device.

05. Verify each NDS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each NDS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

12.02.XX System Level Tests None NuScale US460 SDAA 14.2-39 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-13: Test # 13 Service Air System 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, and N/A N/A all functionality is supported through supported systems testing.

13.00.XX Prerequisites

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

13.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each SAS remotely 1) Operate each valve from the MCR 1) MCR display and local, visual operated valve can be operated and local control panel (if design observation indicate each valve remotely. has local valve control) fully opens and fully closes.

02. Verify each SAS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each SAS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify each SAS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each SAS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

13.02.XX System Level Tests None NuScale US460 SDAA 14.2-40 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-14: Test # 14 Instrument Air System 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.

14.00.XX Prerequisites

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

02. Verify performance testing of air compressor skids have been completed by the manufacturer or a site acceptance test is completed in accordance with manufacturer instructions.

NOTE: Component Level Test 14.01.05 is only required to be tested with the first module.

14.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each IAS remotely-operated 1) Operate each valve from the MCR 1) MCR display and local, visual valve can be operated remotely. and local control panel (if design observation indicate each valve has local valve control). fully opens and fully closes.

02. Verify each IAS air-operated valve Place each valve in its non-safe 1) MCR display and local, visual fails to its safe position on loss of position. observation indicate each valve air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each IAS air-operated valve Place each valve in its non-safe 1) MCR display and local, visual fails to its safe position on loss of position. observation indicate each valve electrical power to its solenoid. 1) Isolate electrical power to each fails to its safe position.

air-operated valve.

04. Verify each IAS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each IAS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

05. Verify the speed of each IAS Align the system to provide a flow path 1) MCR display indicate the speed of variable-speed compressor can to operate a selected compressor. each compressor obtains both be manually controlled. 1) Vary the compressor speed from minimum and maximum minimum to maximum from the compressor speeds.

MCR.

14.02.XX System Level Tests None NuScale US460 SDAA 14.2-41 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-15: Test # 15 Control Room Habitability System 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 CRB by nonsafety-related 15.02.01 providing clean breathing air to the 15.02.02 CRE and maintaining a positive 15.02.03 control room pressure during high radiation or loss of offsite power conditions.

2. The CRHS supports the CRB by nonsafety-related 15.02.01 providing high pressure, clean 15.02.02 breathing air in air bottles for use.

3. The CRVS supports the CRB by nonsafety-related 15.02.01 providing isolation of the CRE from the surrounding areas and outside environment via isolation dampers.

4. The plant protection system (PPS) nonsafety-related 15.02.01 supports the CRHS by providing actuation and control signals.

5. The CRVS supports the PPS by nonsafety-related 15.02.01 providing instrument information signals relating to isolation of the CRE and activation of the CRH system.

6. The CRVS supports the CRB by nonsafety-related 15.02.01 (radiation detection) isolating the CRVS outside air 16.02.03 (smoke/toxic gas) intake from the environment and operating CRVS in recirculation mode to prevent exposure to smoke and toxic gas, or when radiation is detected downstream of the charcoal filtration unit.

7. The PPS supports the CRVS by nonsafety-related 15.02.01 providing actuation and control signals to the CRE isolation dampers.

15.00.XX Prerequisites

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

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

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

04. Component Level Tests 01. and 02. must be performed under preoperational test conditions that approximate design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

NuScale US460 SDAA 14.2-42 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-15: Test # 15 Control Room Habitability System (Continued) 15.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each CRHS Place the CRHS air bottles in service. 1) MCR workstation display, safety remotely-operated valve can be Place CRVS in service to supply air to display instrument display and operated remotely. the CRE. local, visual observation indicate

1) Operate each valve from the MCR. each valve fully opens and fully closes under preoperational temperature, differential pressure, and flow conditions.

[ITAAC 03.01.02]

02. Verify each CRHS Place the CRHS air bottles in service. 1) MCR display, safety display solenoid-operated valve fails to its Place CRVS in service to supply air to instrument display and local, safe position on loss of electrical the CRE. visual observation indicate each power to its solenoid. 1) Place each valve in its non-safe valve fails to its safe position position. Isolate electrical power to under preoperational temperature, its solenoid. differential pressure, and flow conditions.

[ITAAC 03.01.03]

03. Verify each CRHS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each CRHS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

15.02.XX System Level Tests 15.02.01 Test Objective Test Method Acceptance Criteria

1. Verify PPS provides actuation Place the CRVS in automatic 1) MCR workstation display and signals for CRHS and CRVS. operation. local, visual observation indicate

2. CRHS realigns to provide Place the CRHS air bottles in service. the following:

breathable air to the CRE under Place CRVS in service to supply air to a. The CRVS filter unit bypass, inlet, accident conditions. the CRE. and outlet dampers close.

3. CRVS realigns to isolate outside Start CRVS filter unit. b. The CRVS filter unit fan stops.

air dampers and CRE under 1) Initiate each of the following real or c. The CRVS control room envelope accident conditions. simulated CRHS actuation signals: isolation dampers close.

a. High radiation signal downstream d. The CRHS air supply isolation of the CRVS filter unit. valves open.

b. Loss of AC power. e. CRHS pressure relief isolation valves open.

f. CRVS filtration unit stops.

g. CRVS general exhaust fan stops.

h. CRVS battery room exhaust fans stop.

[ITAAC 03.09.02]

(items 1.a. through 1.h.)

2) PPS generates alarms in the MCR for the following:

a. High radiation.

b. Loss of AC power.

NuScale US460 SDAA 14.2-43 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-15: Test # 15 Control Room Habitability System (Continued) 15.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify emergency pressurized air Align air bottles for testing. Assume 25 1) bottles have sufficient volume to percent of the bottles are unavailable a. The CRE described in provide 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of breathable air and use 1/6 of the remaining bottles to Section 6.4.2 maintains a positive through both the main and backup simulate a test conduct of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (12 pressure relative to the adjacent supply flow path to the CRE hours/72 hours). areas as specified in Table 6.4-1 described in Section 6.4.2. 1) Initiate a real or simulated CRHS as indicated by the CRE actuation signal to isolate the CRE. differential pressure 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]

2) 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 for main supply flow path and align the the main flow path is maintained manual backup flow path to the as specified in Table 6.4-1 for the CRE. Align air bottles for testing. duration of the test.

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

15.02.03.

Test Objective Test Method Acceptance Criteria

1. The air exfiltration from the CRE 1) Perform an air exfiltration test of 1) The measured air exfiltration flow does not exceed the air exfiltration the CRE at 1/8 in. wg. of positive rate does not exceed the unfiltered flow rate identified in the CRHS pressure with respect to inleakage flow rate assumed in the exfiltration/infiltration analysis. surrounding areas by performing dose analysis identified in tracer gas testing in accordance Table 6.4-1.

with ASTM E741. [ITAAC 03.01.01]

NuScale US460 SDAA 14.2-44 Revision 0

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

The CRVS is described in Sections 6.4.3 and 9.4.1 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 16.02.01 providing cooling, heating and 16.02.02 humidity control to maintain a suitable environment for the safety and comfort of plant personnel.

2. The CRVS supports the systems nonsafety-related 16.02.01 located in the CRB by providing 16.02.02 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 16.02.03(smoke/toxic gas) isolating the CRVS outside air 15.02.01 (radiation) intake from the environment and operating the CRVS in recirculation mode to prevent exposure to smoke and toxic gas, or when radiation is detected downstream of the charcoal filtration unit.

4. The CRVS supports the CRB by nonsafety-related 16.02.01(CRB positive pressure) maintaining the CRB at a positive 17.02.01(RXB negative pressure) ambient pressure relative to the 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 16.02.03 providing actuation and control signals to the outside air isolation dampers.

6. The CRVS supports the CRB by nonsafety-related 16.02.04 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:

1. The CRVS supports the CRB by nonsafety-related 15.02.01 isolating the CRVS outside air intake when radiation is detected downstream of the charcoal filtration unit.

2. The CRVS supports the CRB by nonsafety-related 15.02.01 providing isolation of the CRE from the surrounding areas and outside environment via isolation dampers.

NuScale US460 SDAA 14.2-45 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-16: Test # 16 Normal Control Room HVAC System (Continued)

3. The CRVS supports the PPS by nonsafety-related 15.02.01 providing instrument information signals relating to isolation of the CRE and activation of the CRHS.

16.00.XX Prerequisites

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

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

03. 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.]

04. 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.]

05. Component Level Test 16.01.07 must be performed under preoperational test conditions that approximate design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

06. Verify CRVS air filtration unit heater testing specified in RG 1.140 C.4.g is completed and the test records have been approved.

16.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each CRVS 1) Operate each damper from the 1) MCR display and local, visual remotely-operated damper can be MCR and local control panel (if observation indicate each damper operated remotely. design has local damper control). fully opens and fully closes.

02. Verify CRVS dampers 1) Open each damper actuated by a 1) MCR display and local, visual automatically close on associated smoke or fire signal. Initiate an observation indicate each damper smoke or fire signals. alarm signal for each damper. closes.

03. Verify each required CRVS fan 1) Initiate an alarm signal for each 1) MCR display and local, visual stops on actuation of its fan. observation indicate each fan associated fire or smoke alarm. stops.

04. Verify each CRVS pressurization 1) Initiate an alarm signal for each 1) MCR display and local, visual fan starts automatically on the fan. observation indicate each actuation of its associated fire or pressurization fan starts.

smoke alarm.

05. Verify the fan speed of each 1) Vary the speed of each fan from 1) MCR display indicates the speed CRVS variable-speed fan can be the MCR and local control panel (if of each fan varies from minimum manually controlled. design has local fan control). to maximum fan speed.

06. Verify the standby CRVS main 1) Place an AHU in service. Place the 1) MCR display and local, visual supply AHU starts automatically standby AHU in automatic control. observation indicate the standby on the stop of the operating CRVS Stop the operating AHU. AHU starts.

main supply AHU.

07. Verify each CRVS CRE isolation 1) Place each damper in its non-safe 1) Each CRVS CRE isolation damper damper fails to its safe position on position. Isolate electrical power to fails to its closed position on loss loss of electrical power to its its solenoid. of electrical power under solenoid. preoperational temperature, differential pressure, and flow conditions while the CRVS is supplying flow to the CRE.

[ITAAC 03.02.01]

08. Verify each CRVS remotely 1) Operate each fan from the MCR 1) MCR display and local, visual operated fan can be operated and local control panel (if design observation indicate each fan remotely. has local fan control). starts and stops.

NuScale US460 SDAA 14.2-46 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-16: Test # 16 Normal Control Room HVAC System (Continued)

09. Verify each CRVS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each CRVS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

16.02.XX System Level Tests 16.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify CRB design temperatures Place the CRVS in automatic 1) The temperature and humidity of and humidity monitored by the operation. rooms and areas monitored by the MCR are maintained at design 1) Record the CRB temperatures and MCR satisfy the design temperature and humidity humidity indications monitored by temperature and humidity conditions during normal the MCR. requirements.

operation. 2) Measure the CRB pressure relative 2) The CRVS maintains a positive

2. Verify The CRVS maintains a the outside environment. pressure of greater than or equal positive pressure in the CRB 3) Measure the air flow rate to the to 0.125 inches water gauge in the relative to the outside environment battery rooms. CRB relative to the outside while the CRVS is operating in environment, while operating in normal alignment. the normal operating alignment.

3. Verify the CRVS maintains the air [ITAAC 03.02.02]

flow to the battery rooms to 3) Measured flow to the battery maintain hydrogen concentration rooms is equal to or greater than to less than 1 percent by volume. the flow specified by the air flow balance.

[ITAAC 03.02.03]

16.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify CRB design temperatures Align the CHWS standby chiller to cool 1) The temperature and humidity of and humidity monitored by the each CRVS main supply AHU. rooms and areas monitored by the MCR are maintained at design 1) Place the CRVS in automatic MCR satisfy the design temperature and humidity operation. temperature and humidity conditions while cooling to the requirements.

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

16.02.03.

Test Objective Test Method Acceptance Criteria

1. Verify PPS actuates CRVS Place the CRVS in automatic 1) Outside air dampers close to outside air dampers when toxic operation. isolate makeup air.

gas or smoke is detected in the 1) Initiate a simulated high smoke or makeup air ductwork. toxic gas signal for the makeup air ductwork upstream of the CRVS filter unit.

16.02.04.

Test Objective Test Method Acceptance Criteria

1. Verify the CRVS automatically Place the CRVS in automatic 1) Outside air is diverted through the responds to mitigate the operation. CRVS filter unit by closing the consequences of high radiation in 1) Initiate a real or simulated high CRVS filter unit bypass dampers the outside air. radiation signal for the outside air and opening the CRVS filter unit ductwork upstream of the CRVS isolation dampers.

filter unit. 2) The CRVS filter unit fan starts.

[ITAAC 03.09.01]

(items 1 and 2)

NuScale US460 SDAA 14.2-47 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-17: Test # 17 Reactor Building HVAC System 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 17.02.01 providing cooling, heating and 17.02.02 humidity control to maintain a 90.03.01 suitable environment for the safety and comfort of plant personnel.

2. The RBVS supports the systems nonsafety-related 17.02.01 located in the RXB by providing 17.02.02 cooling, heating and humidity 90.03.01 control to maintain a suitable environment for the operation of system components.

3. The RBVS supports the RXB by nonsafety-related 17.02.01 maintaining the RXB at a negative 17.02.03 ambient pressure relative to the outside atmosphere to control the movement of potentially airborne radioactivity from the RXB to the environment.

4. The CRVS supports the CRB by nonsafety-related 17.02.01 (RXB negative pressure) maintaining the CRB at a positive 16.02.01 (CRB positive pressure) ambient pressure relative to the 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 RWBVS supports the RWB nonsafety-related 17.02.03 (off-normal RBVS exhaust by maintaining the RWB at a alignment) negative ambient pressure relative 18.02.01 (normal RBVS exhaust to the outside atmosphere to alignment) control the movement of potentially airborne radioactivity from the RWB to the environment.

17.00.XX Prerequisites

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

02. Verify an RBVS air balance is performed and the RBVS air balance records have been approved. [This prerequisite is not required for component-level tests.] (Note: The RBVS is designed to move air from areas that are not contaminated or are expected to have low levels of contamination to areas that are likely to be more contaminated.)

03.RBVS HEPA and charcoal adsorbers have been installed and tested. [This prerequisite is not required for component-level tests.]

04. Verify spent fuel pool exhaust charcoal and HEPA filter unit heater bank testing specified in RG 1.140 C.4.g is completed and the test records have been approved.

17.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each RBVS 1) Operate each damper from the 1) MCR display and local, visual remotely-operated damper can be MCR and local control panel (if observation indicate each damper operated remotely. design has local damper control). fully opens and fully closes.

NuScale US460 SDAA 14.2-48 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-17: Test # 17 Reactor Building HVAC System (Continued)

02. Verify each RBVS damper fails to Place each damper in its non-safe 1) MCR display and local, visual its safe position on loss of position. observation indicate each damper electrical power, if is designed to 1) Isolate electrical power to the fails to its safe position.

do so. damper.

03. Verify RBVS dampers Open each damper actuated by a 1) MCR display and local, visual automatically close on associated smoke or fire signal. observation indicate each damper smoke or fire signals. 1) Initiate an alarm signal for each closes.

damper.

04. Verify each required RBVS fan 1) Initiate an alarm signal for each 1) MCR display and local, visual stops on actuation of its fan. observation indicate each fan associated fire or smoke alarm. stops.

05. Verify the fan speed of each 1) Vary the speed of each fan from 1) MCR display indicates the speed RBVS variable-speed fan can be the MCR and local control panel (if of each fan varies from minimum manually controlled. design has local fan control). to maximum fan speed.

06. Verify each standby RBVS air 1) Place an AHU in service. Place the 1) MCR display and local, visual handling unit starts automatically standby AHU in automatic control. observation indicate the standby on the stop of the operating RBVS Stop the operating AHU. AHU starts.

AHU.

07. Verify each standby RBVS fan coil 1) Place an FCU in service. Place the 1) MCR display and local, visual unit (FCU) starts automatically on standby FCU in automatic control. observation indicate the standby the stop of the operating RBVS Stop the operating FCU. FCU starts.

fan coil unit.

08. Verify each RBVS 1) Start and stop each fan from the 1) MCR display and local, visual remotely-operated fan can be MCR and local panel (if design has observation indicate each fan started and stopped remotely. local fan control). started and stopped.

09. Verify each RBVS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each RBVS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

17.02.XX System Level Tests 17.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify RXB design temperatures Place the RBVS supply, general area 1) The temperature and humidity of and humidity monitored by the exhaust and spent fuel pool exhaust in rooms and areas monitored by the MCR are maintained at design automatic operation. MCR satisfy the design temperature and humidity Place the RWBVS in automatic temperature and humidity conditions during normal operation. requirements.

operation. 1) Record the RXB temperatures and 2) MCR display indicates the RBVS

2. Verify the RBVS maintains a humidity indications monitored by maintains a negative pressure in negative pressure in the RXB the MCR. the RXB relative to the outside relative to the outside environment 2) Measure the RXB pressure relative environment while operating in the while the RBVS is operating in the outside environment. normal operating alignment.

normal alignment. 3) Measure the RWB pressure [ITAAC 03.03.01]

3. Verify The RBVS maintains a relative the outside environment. 3) MCR display indicates the RBVS negative pressure in the RWB 4) Measure the air flow rate to the maintains a negative pressure in relative to the outside environment battery rooms. the RWB relative to the outside while the RBVS is operating in environment while operating in the normal alignment. normal operating alignment.

4. Verify the RBVS maintains the air [ITAAC 03.03.02]

flow to the battery rooms to 4) Measured flow to the battery maintain hydrogen concentration rooms is equal to or greater than to less than 1 percent by volume. the flow specified by the air flow balance.

[ITAAC 03.03.03]

NuScale US460 SDAA 14.2-49 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-17: Test # 17 Reactor Building HVAC System (Continued) 17.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify design temperatures of the 1) Place the RBVS air handling units 1) The temperature and humidity of following rooms can be controlled with installed direct expansion coils rooms and areas monitored by the using AHUs with installed direct in automatic operation. MCR satisfy the design expansion coils. temperature and humidity

a. I&C equipment rooms requirements.

b. Battery rooms

c. Battery charger rooms 17.02.03.

Test Objective Test Method Acceptance Criteria

1. Verify RBVS automatic alignment Place the RBVS general area exhaust, 1) The RBVS general area exhaust on a simulated spent fuel pool RBVS spent fuel pool exhaust, and isolation dampers for the modules hi-hi radiation level. RWBVS exhaust in automatic and dry dock areas are closed.

2. Verify The RBVS maintains a operation. 2) The RBVS diverts spent fuel pool negative pressure in the RXB 1) Place the RBVS supply in exhaust flow to charcoal relative to the outside environment automatic operation. adsorbers and additional HEPAs while the RBVS is operating in 2) Place the RWBVS supply system in the spent fuel pool charcoal filter accident alignment. in automatic operation. units.

3. Verify The RWBVS maintains a 3) Simulate a Hi-Hi radiation signal in [ITAAC 03.09.03]

negative pressure in the RWB the spent fuel pool exhaust (items 1 and 2) relative to the outside environment upstream of the spent fuel pool 3) The RBVS and SFP exhaust fans while the RBVS is operating in charcoal filter units. speed is adjusted to maintain the accident alignment. design negative pressure in the RXB and RWB relative to the outside environment while the RBVS is operating in the off-normal alignment.

NuScale US460 SDAA 14.2-50 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-18: Test # 18 Radioactive Waste Building HVAC System 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 nonsafety-related 18.02.01 by providing cooling, heating and humidity control to maintain a suitable environment for the safety and comfort of plant personnel.

2. The RWBVS supports the nonsafety-related 18.02.01 systems 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 nonsafety-related 18.02.0 (normal RBVS exhaust by maintaining the RWB at a alignment) negative ambient pressure relative 17.02.03 (off-normal RBVS exhaust to the outside atmosphere to alignment) control the movement of potentially airborne radioactivity from the RWB to the environment.

18.00.XX Prerequisites

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

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

18.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each RWBVS 1) Operate each damper from the 1) MCR display and local, visual remotely-operated damper can be MCR and local control panel (if observation indicate each damper operated remotely. design has local damper control). fully opens and fully closes.

02. Verify each RWBVS damper fails Place each damper in its non-safe 1) MCR display and local, visual to its safe position on loss of position. observation indicate each damper electrical power, if it is designed to 1) Isolate electrical power to the fails to its safe position.

do so. damper.

03. Verify RWBVS dampers Open each damper actuated by a 1) MCR display and local, visual automatically close on associated smoke or fire signal. observation indicate each damper smoke or fire signals. 1) Initiate an alarm signal for each closes.

damper.

04. Verify each required RWBVS fan 1) Initiate an alarm signal for each 1) MCR display and local, visual stops on actuation of its fan. observation indicate each fan associated fire or smoke alarm. stops.

05. Verify the fan speed of each 1) Vary the speed of each fan from 1) MCR display indicates the speed RWBVS variable-speed fan can the MCR and local control panel (if of each fan varies from minimum be manually controlled. design has local fan control). to maximum fan speed.

06. Verify the standby RWBVS main Place an AHU in service. Place the 1) MCR display and local, visual supply AHU starts automatically standby AHU in automatic control. observation indicate the standby on the stop of the operating 1) Stop the operating recirculation AHU starts.

RWBVS main supply AHU. AHU.

07. Verify each standby RWBVS FCU Place an FCU in service. Place the 1) MCR display and local, visual starts automatically on the stop of standby FCU in automatic control. observation indicate the standby the operating RWBVS FCU. 1) Stop the operating FCU. FCU starts.

NuScale US460 SDAA 14.2-51 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-18: Test # 18 Radioactive Waste Building HVAC System (Continued)

08. Verify each RWBVS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each on an MCS or PCS display, or is display. RWBVS transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

18.02.XX System Level Test 18.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the RWB design 1) Place the RWBVS in automatic 1) The temperature and humidity of temperatures and humidity operation. rooms and areas monitored by the monitored by the MCR are 2) Place the RBVS in automatic MCR satisfy the design maintained at design temperature operation. temperature and humidity and humidity conditions during requirements.

normal operation. 2) MCR display indicates the

2. Verify the RWBVS maintains a RWBVS maintains a negative negative pressure in the RWB pressure in the RWB relative to relative to the outside environment the outside environment while while the RWBVS is operating in operating in the normal operating normal alignment. alignment.

NuScale US460 SDAA 14.2-52 Revision 0

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

The 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 #

1. The TBVS supports the systems nonsafety-related 19.02.01 located in the TGB by providing cooling, heating and humidity control to maintain a suitable environment for the operation of system components.

19.00.XX Prerequisites

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

19.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each air conditioning unit 1) Start air conditioning and 1) Air conditioning unit and and condensing unit operates to condensing unit. condensing unit operates to maintain room temperature. maintain room temperature.

02. Verify TBVS dampers Open each damper actuated by a 1) MCR display and local, visual automatically close on associated smoke or fire signal. observation indicate each damper smoke or fire signals. 1) Initiate an alarm signal for each closes.

damper.

03. Verify each required TBVS fan 1) Initiate an alarm signal for each 1) MCR display and local, visual stops on actuation of its fan. observation indicate each fan associated fire or smoke alarm. stops.

04. Verify each TBVS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each TBVS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

19.02.XX System Level Tests 19.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the TGB battery and battery 1) Place the turbine bypass system 1) The temperature and humidity of charger room design battery and battery charger room TGB battery and battery charger temperatures are maintained at ventilation units in automatic rooms satisfy the temperature and design temperature and humidity operation. humidity requirements.

conditions during normal operation.

NuScale US460 SDAA 14.2-53 Revision 0

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

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

System Function System Function Categorization Function Verified by Test #

1. The RWDS supports the RWB by nonsafety-related 20.02.01 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 20.02.01 collecting radioactive waste in drain sumps and tanks and transfers it to the LRWS for processing.

3. The RWDS supports the UHS by nonsafety-related 20.02.02 providing detection and monitoring of leakage through the UHS liner and the dry dock liner.

4. The LRWS supports the RWDS by nonsafety-related 20.02.01 receiving and processing the 30.02.02 effluent from the RWB radioactive waste drain sumps.

5. The LRWS supports the RWDS by nonsafety-related 20.02.01 receiving and processing the 30.02.02 effluent from the RXB radioactive waste drain sumps.

20.00.XX Prerequisites

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

20.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each RWDS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each RWDS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each RWDS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify the RWDS automatically 1) Initiate a real or simulated high 1) MCR display verifies the RWDS to responds to mitigate a release of radiation signal for the RCCWS RCCWS expansion tank isolation radioactivity. water drain tank. valve is closed.

[ITAAC 03.09.10]

05. Verify each RWDS pump can be Align the RWDS to allow for pump 1) MCR display and local, visual started and stopped remotely. operation. observation indicate each pump

1) Start and stop each pump from the starts and stops.

MCR.

06. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from an RWDS grab flow through the grab sampling obtained.

sample device indicated on the device.

RWDS piping and instrumentation diagram.

NuScale US460 SDAA 14.2-54 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-20: Test # 20 Radioactive Waste Drain System (Continued)

07. Verify each RWDS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each RWDS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

20.02.XX System Level Tests 20.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify RWDS pumps start and Align each RWDS sump or tank to MCR displays and local, visual stop automatically and transfer allow water in a selected sump or tank observation verifies the following:

liquid waste to its design location to be pumped to its design location in 1) The first pump starts on HI level in the LRWS. the LRWS (as indicated by the RWDS and transfers water to its design piping and instrumentation diagrams). location in the LRWS.

1) Fill the selected sump or tank until 2) The second (alternate) pump a HI water level is obtained to start starts on HI-HI level.

the first (primary) pump. 3) Both primary and alternate pumps

2) Continue filling the sump or tank stop on LO level.

until a HI-HI level starts the second 4) The primary pump starts on HI (alternate) pump. level.

3) Stop filling the sump or tank and 5) The alternate pump starts on HI-HI allow the primary and alternate level.

pumps to stop on LO level.

4) Change pump controls to make Pump #2 the primary pump and Pump #1 the alternate pump. Refill the sump or tank until the primary pump starts on HI level.

5) Continue filling the sump or tank until a HI-HI level starts the alternate pump.

20.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify each RWDS equipment 1) Fill the selected sump at a rate that 1) PCS data indicate the sump fill drain sump alarms on a fill rate exceeds the PLDS leakage rate rate alarmed at the PLDS leakage that exceeds the PLDS leakage setpoint. rate setpoint.

rate setpoint.

NuScale US460 SDAA 14.2-55 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-21: Test # 21 Balance-of-Plant Drain System Preoperational test is required to be performed to support sequence of construction turnover of the BPDS.

The BPDS 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 BPDS supports the nonsafety-related 21.02.01 condensate polisher 21.02.07 demineralizers, the cooling tower chemical addition system, 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 TGB, the nonsafety-related 21.02.01 two diesel generators, the 21.02.07 auxiliary boiler, the Central Utility Building (CUB), 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 21.02.01 drains by providing a means to 21.02.07 collect, treat, and transfer the waste water to the UWS.

21.00.XX Prerequisites

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

21.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each BPDS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each BPDS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each BPDS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify each BPDS pump can be Align the BPDS to allow for pump 1) MCR display and local, visual started and stopped remotely. operation. observation indicate each pump

1) Start and stop each pump from the starts and stops. Audible and MCR. visible water hammer are not observed when the pump starts.

05. Verify each BPDS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each BPDS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

NuScale US460 SDAA 14.2-56 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-21: Test # 21 Balance-of-Plant Drain System (Continued) 21.02.XX System Level Tests 21.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify BPDS automatically Align each BPDS sump or tank to allow MCR displays and local, visual controlled pumps in sumps and water in a selected sump or tank to be observation verifies the following:

tanks with a fire water removal pumped to its design location. If the 1) The primary pump starts on HI pump start and stop automatically sump fill rate in the following test level and transfers water to its and transfer liquid waste to its method is insufficient for automatic design location in the LRWS or design location. start of the alternate pump or fire UWS.

pump, the primary pump or alternate 2) The alternate pump starts on HI-HI pump may be temporarily removed level.

from service to allow an increase in the 3) The fire water removal pump starts sump level. on HI-HI-HI level.

1) Verify that Pump #1 is set to the 4) The fire water removal pump stops primary pump and Pump #2 is set on LO level.

to alternate. Fill the selected sump 5) Both primary and alternate pumps or tank until a HI water level is stop on LO level.

obtained to start the primary pump. 6) The primary pump starts on HI

2) Continue filling the sump or tank level.

until a HI-HI level starts the 7) The alternate pump starts on HI-HI alternate pump. level.

3) Fill the sump or tank until a HI-HI-HI level starts the fire water removal pump.

4) Stop filling the sump or tank to allow the fire water removal pump to stop on LO level.

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

6) Change pump controls to make Pump #2 the primary pump and Pump #1 the alternate pump, and refill the sump or tank until the primary pump starts on HI level.

7) Continue filling the sump or tank until a HI-HI level starts the alternate pump.

Note: Pump #1 and Pump #2 are not the actual names of the pumps, these names are used to differentiate between the two pumps.

21.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify the BPDS automatically Place a chemical waste water sump 1) The chemical waste water sump responds to mitigate a release of pump in operation. pump stops.

radioactivity. 1) Initiate a real or simulated high 2) Chemical waste collection sump to radiation signal on the 00 CPS BPDS collection tank isolation regeneration skid waste effluent. valve is closed.

Repeat the test for each pump. 3) Chemical waste collection sump to LRWS high-conductivity waste (HCW) tank isolation valve is closed.

[ITAAC 03.09.06]

(items 1 through 3)

NuScale US460 SDAA 14.2-57 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-21: Test # 21 Balance-of-Plant Drain System (Continued) 21.02.03.

Test Objective Test Method Acceptance Criteria

1. Verify the BPDS automatically Place a waste water sump pump in 1) The waste water sump pump responds to mitigate a release of operation. stops.

radioactivity. 1) Initiate a real or simulated high 2) Waste water sump discharge to radiation signal in the BPDS TGB BPDS collection tank isolation floor drains. valve is closed.

Repeat the test for each pump. 3) Waste water sump discharge to LRWS HCW tank isolation valve is closed.

[ITAAC 03.09.06]

(items 1 through 3) 21.02.04.

Test Objective Test Method Acceptance Criteria

1. Verify the BPDS automatically Place a waste water sump pump in 1) The chemical waste water sump responds to mitigate a release of operation. pump stops.

radioactivity. 1) Initiate a real or simulated high 2) Chemical waste collection sump to radiation signal in the BPDS BPDS collection tank isolation auxiliary blowdown cooler valve is closed.

condensate. 3) Chemical waste collection sump to Repeat the test for each pump. LRWS HCW tank isolation valve is closed.

[ITAAC 03.09.06]

(items 1 through 3) 21.02.05.

Test Objective Test Method Acceptance Criteria

1. Verify BPDS automatically Align each BPDS sump or tank to allow MCR displays and local, visual controlled pumps, in sumps and water in a selected sump or tank to be observation verifies the following:

tanks without a fire water removal pumped to its design location. 1) The primary pump transfers water pump, stop automatically and 1) Verify that Pump #1 is set to the to its design location in the LRWS transfer liquid waste to its design primary pump and Pump #2 is set or UWS.

location. to alternate. Fill the selected sump 2) Both primary and alternate pumps or tank until a HI water level alarm stop on LO level.

is obtained and start the primary pump.

2) Continue filling the sump or tank until a HI-HI level alarm and start the alternate pump.

3) Stop filling the sump or tank to allow the primary and alternate pumps to stop on LO level.

Note: Pump #1 and Pump #2 are not the actual names of the pumps; these names are used to differentiate between the two pumps.

NuScale US460 SDAA 14.2-58 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-22: Test # 22 Fire Protection System Preoperational test is required to be performed once, and is conducted in accordance with the applicable criteria in codes and standards listed in Table 9.5.1-1.

The 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

site plant cooling structures

CUB

CRB

high voltage AC electrical distribution 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 tests providing audible and visual alarms to alert operators in the MCR.

22.00.XX Prerequisites

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

02. Verify a pump curve test is completed for the fire protection pumps.

22.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify position indication for each 1) Operate each valve manually. 1) MCR display and local, visual FPS manual valve with remote observation indicate each valve position indication. fully opens and fully closes.

02. Verify each FPS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each FPS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

NuScale US460 SDAA 14.2-59 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-22: Test # 22 Fire Protection System (Continued)

04. Verify each FPS pump can be Align the FPS to allow for pump 1) MCR display and local, visual started and stopped. operation. observation indicate each pump

1) Start each pump locally. starts. Audible and visible water

2) Stop each pump locally. hammer are not observed when the pump starts.

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

05. Verify automatic operation of FPS 1) 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. 1) The jockey pump maintains the

2) Stop the jockey pump and simulate FPS header greater than or equal a low FPS header pressure to start to 10 psig above the pressure the electric fire pump. setting for the automatic start of

3) Stop the electric fire pump and the electric fire pump.

simulate a low FPS header 2) The electric fire pump starts.

pressure to start the diesel fire Audible and visible water hammer pump. are not observed when the pump starts.

3) The diesel pump starts. Audible and visible water hammer are not observed when the pump starts.

06. Verify each valve with a tamper 1) Partially close each FPS manual 1) An alarm is received in the MCR switch alarms when partially valve with a tamper switch to its when each valve is partially closed. alarm position (approximately 20 closed.

percent of its total travel distance).

07. Verify each smoke and fire Isolate the water supply to each 1) The MCR receives an alarm and detector provides audible and preaction or deluge sprinkler before indication from each smoke and visual alarms and annunciation in performing this test to prevent wetting fire detector.

the MCR. equipment.

1) Simulate a smoke or fire signal to each detector.

08. Verify fire pump flow meets its fire Align the FPS for pump operation 1) The electric fire pump meets its protection volumetric flow rate. through the recirculation line. design volumetric flow rate.

1) Start the electric fire pump. 2) The diesel fire pump meets its

2) Start the diesel fire pump. design volumetric flow rate.

[ITAAC 03.07.02]

(items 1 and 2)

09. Verify each suppression system 1) Isolate fluid source and simulate 1) Valve opens and send alarm to actuation valve opens and alarms actuation signal. MCR.

in the MCR when signal is received.

10. Verify each suppression system 1) Isolate fluid source and simulate 1) Flow alarm sent to MCR.

flow switch alarms in the MCR. flow signal.

11. Verify each FPS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each FPS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

22.02.XX System Level Tests None NuScale US460 SDAA 14.2-60 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-23: Test # 23 Fire Detection System Preoperational test is required to be performed once, and is conducted in accordance with the applicable criteria in codes and standards listed in Table 9.5.1-1.

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 nonsafety-related As described in Test Abstract Table 14.2-22 Table 14.2-22 01.00.XX Prerequisites As described in Test Abstract Table 14.2-22 23.00.XX Component Level Tests As described in Test Abstract Table 14.2-22 23.01.XX System Level Tests As described in Test Abstract Table 14.2-22 NuScale US460 SDAA 14.2-61 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-24: Test # 24 Main Steam System Preoperational test is required to be performed for each NPM.

The MSS is described in Section 10.3. MSS 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 TGS by nonsafety-related 29.02.02 providing steam to the TGS. 94.03.01

2. The MSS supports the nonsafety-related 56.02.04 containment system (CNTS) by providing secondary isolation of the main steam (MS) lines.

3. The MSS supports the DHRS by nonsafety-related 56.02.01 providing a backup means for required boundary conditions for DHRS operation.

24.00.XX Prerequisites

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

24.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each MSS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each MSS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each MSS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

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

main turbine. 1) FWH high level 1) Extraction steam block valve

2) Turbine trip closes.

2) Extraction steam non-return check valve closes.

05. Verify the MSS automatically Initiate a real or simulated high 1) MCR display verifies the following responds to mitigate a release of radiation signal for each of the valves are closed:

radioactivity. following: a. main steam common steam

1) SG #1 main steam line radiation header drain pot control valve

2) SG #2 main steam line radiation b. SG #1 drain pot control valve

c. SG #2 drain pot control valve

[ITAAC 02.07.04]

06. Verify each MSS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each MSS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

24.02.XX System Level Tests None NuScale US460 SDAA 14.2-62 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-25: Test # 25 Condensate and Feedwater System Preoperational test is required to be performed for each NPM.

The FWS is described in Section 10.4.6; 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 CPS by nonsafety-related 27.02.01 providing water for CPS rinse and CPS resin transfer.

2. The FWS supports the TGS by nonsafety-related 29.02.01 cooling superheated steam in the 94.03.01 gland steam desuperheater before the steam entering the gland seals.

3. The FWS supports the CNTS by nonsafety-related 29.02.01 supplying FW to the SGs. 94.03.01

4. The FWS supports the TGS by nonsafety-related 29.02.01 cooling superheated turbine 97.03.01 bypass steam in the turbine bypass desuperheater before the steam entering the main condenser.

5. The FWS supports the CNTS by nonsafety-related 56.02.04 providing secondary isolation of the FW lines.

6. The FWS supports the DHRS by nonsafety-related 56.02.04 providing secondary isolation of the FW lines, ensuring required boundary conditions for DHRS operation.

25.00.XX Prerequisites

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

02. Verify a pump curve test is completed for the FWS pumps.

25.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each FWS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each FWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each FWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify each FWS condensate Align the FWS to allow for pump 1) MCR display and local, visual pump can be started and stopped operation. observation indicate each pump remotely. 1) Start and Stop each pump from the starts and stops. Audible and MCR. visible water hammer are not observed when the pump starts.

05. Verify the pump speed of each Align the FWS to provide a flow path to 1) MCR display indicates the speed FWS variable-speed pump can be operate a selected FWS of each variable speed pump manually controlled. variable-speed pump. obtains both minimum and

1) From the MCR, vary FWS pump maximum pump speeds. Audible speed from minimum to maximum and visible water hammer are not for each FWS pump. observed when the pump starts.

NuScale US460 SDAA 14.2-63 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-25: Test # 25 Condensate and Feedwater System (Continued)

06. Verify condensate pump low flow 1) Align the FWS for automatic short MCR displays and local, visual protection and short cycle cycle cleanup. Place a condensate observation verifies the following:

automatic operation. pump in operation. 1) The short cycle flow is

2) Manually throttle a valve in the automatically maintained by the pump flow path until the flow rate short cycle cleanup flow control reaches the pump minimum flow valve.

setpoint. 2) The condensate pump minimum

3) Open the throttled valve. flow valve is open.

3) The condensate pump minimum flow valve is closed.

07. Verify FW pump low flow 1) Align the FWS for automatic long MCR displays and local, visual protection. cycle cleanup. Place a condensate observation verifies the following:

pump in operation. 1) The long cycle flow is

2) Manually throttle a valve in the automatically maintained by the pump flow path until the flow rate long cycle cleanup flow control reaches the FW pump minimum valve.

flow setpoint. 2) The FW pump minimum flow valve

3) Open the throttled valve. is open.

3) The FW pump minimum flow valve is closed.

08. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from an FWS grab flow through the grab sampling obtained.

sample device. device.

09. Verify each FWS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each FWS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

25.02.XX System Level Tests None NuScale US460 SDAA 14.2-64 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-26: Test # 26 Feedwater Treatment System Preoperational test is required to be performed once each for the shared or common components. The module-specific portions of the test must be completed once for each NPM.

The FWTS is described in Section 10.4.8 and the function verified by this test and power ascension testing is:

System Function System Function Categorization Function Verified by Test #

1. The FWTS supports the FWS by nonsafety-related Component-level tests controlling and maintaining FW 72.03.01 chemistry.

26.00.XX Prerequisites

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

26.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each FWTS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each FWTS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each FWTS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify each FWTS pump can be Align the FWTS to allow for pump 1) MCR display and local, visual started and stopped remotely and operation. observation indicate each pump locally (if designed). 1) Start and stop each starts and stops. Audible and remotely-controlled pump from the visible water hammer are not MCR. observed when the pump starts.

2) Start and stop each locally-controlled pump locally.

05. Verify the speed of each FWTS 1) Vary the speed of each pump from 1) MCR display indicates pump variable-speed pump can be the MCR and local control panel (if speed varies from minimum to manually controlled. design has local pump control). maximum pump speed.

06. Verify each FWTS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each FWTS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

26.02.XX System Level Tests None NuScale US460 SDAA 14.2-65 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-27: Test # 27 Condensate Polisher Resin Regeneration System Preoperational test is required to be performed once.

The CPS is described in Section 10.4.5. 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 27.02.01 regenerating the resin that purifies the condensate.

2. The FWS supports the CPS by nonsafety-related 27.02.01 providing water for CPS rinse and CPS resin transfer.

3. The ABS supports the CPS by nonsafety-related Component Level Tests supplying chemical supply piping and connections for neutralization.

27.00.XX Prerequisites:

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

02. NOTE: Component Level Tests may be performed as SAT on vendor supplied skids.

27.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each CPS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely, if not has local valve control). fully opens and fully closes.

performed as part of SAT.

02. Verify each CPS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air, if not performed as part 1) Isolate and vent air to the valve. fails to its safe position.

of SAT.

03. Verify each CPS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid, if not performed as part air-operated valve.

of SAT.

04. Verify each CPS pump can be Align the CPS to allow for pump 1) MCR display and local, visual started and stopped remotely and operation. observation indicate each pump locally (if designed), if not 1) Start and stop each starts and stops. Audible and performed as part of SAT. remotely-controlled pump from the visible water hammer are not MCR. observed when the pump starts.

2) Start and stop each locally-controlled pump locally.

05. Verify the speed of each CPS 1) Vary the speed of each pump from 1) MCR display indicates pump variable-speed pump can be the MCR and local control panel (if speed varies from minimum to manually controlled, if not design has local pump control). maximum pump speed.

performed as part of SAT.

06. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from a CPS grab sample flow through the grab sampling obtained.

device, if not performed as part of device.

SAT.

07. Verify each CPS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each CPS on an MCS or PCS display, or is display, if not performed as part of transmitter. recorded by the applicable control SAT. system historian.

(Test not required if the instrument calibration verified the MCS or PCS display.)

NuScale US460 SDAA 14.2-66 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-27: Test # 27 Condensate Polisher Resin Regeneration System (Continued) 27.02.XX System Level Test 27.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the CPS automatically Align the FWS to support CPS resin 1) The resin transferred to the completes resin regeneration. regeneration. regeneration skid.

Align the ABS to support CPS resin 2) The CPS regeneration cycle regeneration. completed successfully.

1) Automatically transfer the test resin 3) The resin transferred to a bed from a condensate polisher to condensate polisher.

the CPS regeneration skid. 4) ABS steam maintains hot water

2) Initiate an automatic regeneration heater outlet temperature at of the resin. design setpoint during resin

3) Automatically transfer the test resin regeneration.

bed from the CPS regeneration skid to a condensate polisher.

NuScale US460 SDAA 14.2-67 Revision 0

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

The feedwater heater vents and drains system (HVDS) is described in Section 10.4.6. and the functions verified by this test and power ascension testing are:

System Function System Function Categorization Function Verified by Test #

1. The HVDS supports the FWS by nonsafety-related Component level tests venting the FWHs.

2. The HVDS supports the FWS by nonsafety-related Component level tests controlling level in the shell side 94.03.01 FWHs.

28.00.XX Prerequisites

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

28.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each HVDS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each HVDS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each HVDS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify automatic operation of 1) Initiate a simulated turbine trip. Any remote display or local verification HVDS valves to protect the turbine indicates the following:

on turbine trip. 1) Low, intermediate, and high pressure FWH extraction steam supply valves are closed.

2) Low, intermediate, and high pressure FWH air assisted check valves are closed.

3) Low, intermediate, and high pressure FWHs extraction steam dump valves are open.

05. Verify automatic operation of Initiate a simulated signal for the Any remote display or local verification HVDS valves to protect the turbine following system conditions. indicates the following:

on high FWH level. 1) Low pressure FWH high level. 1) Low pressure FWH extraction

2) Intermediate pressure FWH high steam supply valve and low level. pressure FWH extraction steam

3) High pressure FWH high level. dump valve are open.

2) Intermediate pressure FWH extraction steam supply valve and intermediate pressure FWH extraction steam dump valve are open.

3) High pressure FWH extraction steam supply valve and high pressure FWH extraction steam dump valve are open.

NuScale US460 SDAA 14.2-68 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-28: Test # 28 Feedwater Heater Vents and Drains System (Continued)

06. Verify each HVDS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each HVDS on an MCS or PCS display, or is display. system transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

28.02.XX System Level Tests None NuScale US460 SDAA 14.2-69 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-29: Test # 29 Turbine Generator System 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 and power ascension testing are:

System Function System Function Categorization Function Verified by Test #

1. The TGS supports the MSS by nonsafety-related 29.02.01 providing steam bypass from the 97.03.01 MSS to the main condenser.

2. The module heatup system (MHS) nonsafety-related 29.02.01 supports the CVCS by adding heat to primary coolant.

3. The CVCS supports the RCS by nonsafety-related 29.02.01 heating primary coolant.

4. The FWS supports the CNTS by nonsafety-related 29.02.01 supplying FW to the SGs. 93.03.01

5. The FWS supports the TGS by nonsafety-related 29.02.01 cooling superheated turbine 97.03.01 bypass steam in the turbine bypass desuperheater before the steam entering the main condenser.

6. The FWS supports the TGS by nonsafety-related 29.02.01 cooling superheated steam in the 94.03.01 gland steam desuperheater before the steam entering the gland seals.

7. The CVCS supports the ECCS nonsafety-related 29.02.01 valves by providing water to reset the ECCS valves.

8. The MSS supports the TGS by nonsafety-related 29.02.01 providing steam to the TGS. 94.03.01 29.00.XX Prerequisites

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

The following prerequisites are only required for the System Level Tests:

02. Verify Test 07.02.01 is completed to verify the CARS can maintain main condenser vacuum pressure (reference test Table 14.2-7).

03. The SG FW flush is complete.

04. The CARS is automatically maintaining main condenser vacuum.

05. 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.

06. The NPM and supporting systems are aligned to increase RCS temperature and pressure.

29.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each TGS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each TGS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each TGS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

NuScale US460 SDAA 14.2-70 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-29: Test # 29 Turbine Generator System (Continued)

04. Verify each TGS lube oil pump Align the TGS to allow for main lube 1) MCR display and local, visual can be started and stopped oil, auxiliary lube oil, and emergency observation indicate each pump remotely. pump operation. starts and stops. Audible and

1) Start and stop each pump from the visible water hammer are not MCR. observed when the pump starts.

05. Verify the TGS exhaust hood is 1) Initiate a simulated high exhaust 1) Any remote display or the local, protected against high hood temperature. visual observation indicates the temperature. exhaust hood spray valve is open.

06. Verify TGS lubricating oil flow Align the TGS to allow for main lube oil 1) MCR display and local, visual capability by automatic start of the and auxiliary lube oil pump operation. observation indicate the auxiliary auxiliary lube oil pump. Place the TGS main oil pump in normal oil pump starts. Audible and visible service. Place the auxiliary oil pump in water hammer are not observed standby. when the pump starts.

1) Simulate a TGS auxiliary oil pump start.

07. Verify TGS lubricating oil flow Align the TGS to allow for auxiliary 1) MCR displays and local, visual capability by automatic start of the lube oil pump and emergency lube oil observation indicate the TGS emergency direct current (DC) pump operation. Place the turbine emergency oil pump starts.

lube oil pump. generator auxiliary oil pump in normal Audible and visible water hammer service. are not observed when the pump

1) Simulate a turbine generator starts.

emergency oil pump start or simulate a loss of AC power.

08. Verify the turbine stop valve and 1) 1) turbine control valves close on a. Simulate an overspeed trip signal a. The turbine stop valve and turbine turbine overspeed. from the turbine overspeed control valves close.

emergency trip system. b. Each turbine stop valve and

b. Record the stroke times of the turbine control valve close stroke turbine stop valve and the turbine time is within design limits.

control valves. 2)

2) a. The turbine stop valve and turbine

a. Simulate an overspeed trip signal control valves close.

from the governor overspeed b. Each turbine stop valve and detection circuit. turbine control valve close stroke

b. Record the stroke times of the time is within design limits.

turbine stop valve and the turbine control valves.

09. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is obtained.

obtained from the gland seal flow through the grab sample exhauster discharge grab sample device.

device.

10. Verify the turbine can be manually 1) Manually trip the turbine from an 1) The turbine stop valve and turbine tripped. operator workstation in the MCR. control valves close.

11. Verify each TGS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each TGS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

NuScale US460 SDAA 14.2-71 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-29: Test # 29 Turbine Generator System (Continued) 29.02.XX System Level Tests 29.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the CVCS is capable of 1) Close the ECCS valves. 1) CVCS pressure is sufficient as supplying water at sufficient 2) Align the plant to cool the RCS via indicated by closure of the ECCS pressure to close the ECCS the TGS bypass system. valves.

valves. 3) Warm MS lines. 2)

2. Verify the MHS is capable of 4) Place the TGS steam bypass valve a. CVCS supply remains in a heating the RCS to a temperature in automatic control. sub-cooled state while heating the sufficient to obtain criticality. 5) Place the FW regulating valve in RCS using the MHS as verified by

3. Verify the MHS is capable of SG inventory control. CVCS temperature and pressure.

heating the RCS to establish 6) Place the MHS and the CVCS in b. RCS temperature is sufficient to natural circulation flow sufficient to automatic control to heat the RCS. obtain criticality.

obtain criticality. 7) Align the FWS to cool the gland 3) RCS natural circulation flow is

4. Verify the TGS automatically seal steam desuperheater. sufficient to obtain criticality.

controls turbine bypass flow to the 4) The TGS bypass flow maintains main condenser. steam pressure at setpoint.

5. Verify the FWS automatically 5) The FW flow to the SG is controls flow to the SGs to maintained at setpoint.

maintain SG inventory. 6) The cooled TGS bypass

6. Verify the FWS automatically temperature is maintained at cools the TGS bypass steam in setpoint.

the MS desuperheater. 7) A local grab sample is successfully

7. Verify a local grab sample can be obtained at RCS normal operating obtained from an MHS grab pressure and maximum sample device. temperature achievable.

8. Verify the FWS automatically 8) The cooled gland seal steam cools the TGS gland steam in the temperature is maintained at gland steam desuperheater. setpoint.

9. Verify CCT level is automatically 9) CCT level is maintained at setpoint controlled while receiving bypass while receiving bypass steam.

steam. 10) Water hammer indications:

10. Verify no dynamic effects caused a. Audible indications of water by changes in fluid flow. hammer are not observed.

b. No damage to pipe supports or restraints.

c. No damage to equipment.

d. No equipment leakage 29.02.02.

This test may be performed after the completion of Test 29.02.01 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

1. Verify the maximum main turbine Place the main turbine in service as 1) The maximum main turbine speed speed that can be obtained using follows: is obtained.

the MHS to heat the RCS. 1) Ensure the RCS is at normal operating pressure and at maximum temperature achievable by warming the RCS using MHS heating.

2) Place turbine on turning gear with seal steam in service.

3) Warm up turbine to required temperature.

4) Increase main turbine speed.

NuScale US460 SDAA 14.2-72 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-30: Test # 30 Liquid Radioactive Waste System Preoperational test is required to be performed once.

The LRWS is described in Section 11.2 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 30.02.02 radioactive waste system (SRWS) Component-level test 30.01.11 by receiving and processing liquid 32.02.07 radioactive waste from the SRWS dewatering skid.

2. The LRWS supports the PCWS by nonsafety-related 30.02.02 receiving contaminated pool water Component-level tests to aid in the removal of titrated water or boron. Treated liquid radioactive waste has the option to return to the pool as makeup.

3. The LRWS supports the CVCS by nonsafety-related 30.02.02 receiving and processing primary 33.02.01 coolant from CVCS letdown.

4. The LRWS supports the RWDS by nonsafety-related 30.02.02 receiving and processing the 20.02.01 effluent from the RWB radioactive waste drain sumps.

5. The LRWS supports the RWDS by nonsafety-related 30.02.02 receiving and processing the 20.02.01 effluent from the RXB radioactive waste drain sumps.

6. LRWS supports the CVCS by nonsafety-related 30.02.01 receiving and processing the noncondensable gases and vapor from the PZR.

7. LRWS supports the PCWS by nonsafety-related 30.02.02 processing any fluid collected in Component-level tests the drain sump of the secondary containment tank.

8. The NDS supports the LRWS by nonsafety-related 30.02.01 providing nitrogen for purging of Table 14.2-12 component-level tests the LRWS.

NuScale US460 SDAA 14.2-73 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-30: Test # 30 Liquid Radioactive Waste System (Continued) 30.00.XX Prerequisites

01. Required ANSI/ANS-55.6 construction testing is completed.

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

30.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each LRWS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control) fully opens and fully closes.

02. Verify each LRWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each LRWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify each LRWS pump can be Align the LRWS to allow for pump 1) MCR display and local, visual started and stopped remotely. operation. observation indicate each pump

1) Start and stop each pump from the starts and stops. Audible and MCR. visible water hammer are not observed when the pump starts.

05. Verify the speed of each LRWS Align the LRWS to provide a flow path 1) MCR display indicates the speed variable-speed pump can be to operate a selected pump. of each obtains both minimum and manually controlled. 1) Vary the LRWS pump speed from maximum pump speeds.

minimum to maximum from the MCR.

06. Verify LRWS 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, 1) LRWS discharge to the utility water 1) The LRWS discharge to the utility low dilution flow and underground discharge basin high radiation water discharge basin isolation pipe break. signal. valves close.

2) LRWS discharge to the utility water [ITAAC 03.09.07]

discharge basin low dilution flow 2) The LRWS discharge to the utility signal. water discharge basin isolation

3) LRWS discharge to the utility water valves close.

discharge basin low guard pipe 3) The LRWS discharge to the utility pressure signal. water discharge basin isolation valves close.

07. Verify the LWRS automatically 1) Initiate a simulated high area 1) MCR display verifies the LRWS to responds to mitigate a release of radiation signal for the GRWS GRWS discharge isolation valve is radioactivity. charcoal bed cubicle. closed.

[ITAAC 03.09.07]

08. Verify tank valves operate to Simulate an inservice tank high level 1) MCR display and local, visual ensure uninterrupted waste signal for each of the following tanks: observation indicate the inservice receiving. 1) Low-conductivity waste (LCW) tank fill valve is closed and the collection tank A and B. standby tank fill valve is open.

2) HCW collection tank A and B.

3) LCW sample tank A and B HCW sample tank A and B.

09. Verify degasifier valves operate to 1) Initiate a simulated high degasifier 1) MCR display and local, visual ensure uninterrupted waste level signal. observation indicate the inservice receiving. 2) Initiate a simulated high degasifier degasifier fill valve is closed and pressure signal. the standby degasifier fill valve is open.

NuScale US460 SDAA 14.2-74 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-30: Test # 30 Liquid Radioactive Waste System (Continued)

10. Verify LRWS pumps automatically Align the LRWS to allow each of the MCR displays and local, visual operate to prevent tank overflow. following LRWS transfer pumps to observation indicate the following:

automatically transfer effluent to one of 1) 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 2) The transfer pump stops.

and 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.

1) Simulate a HI-HI level signal in each of the above tanks.

2) Simulate a low level signal in each of the above tanks.

11. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from a LRWS grab flow through the grab sampling obtained.

sample device indicated on the device.

LRWS piping and instrumentation diagram.

12. Verify SRWS dewatering skid Align SRWS dewatering skid discharge 1) SRWS dewatering skid effluent is effluent can be transferred to to one of the LRWS HCW collection transferred to the LRWS LRWS HCW collection tanks. tanks. high-conductivity waste collection Fill the SRWS dewatering skid HIC to tank.

above the low level pump stop 2) The SRWS dewatering skid setpoint. diaphragm pump is stopped.

1) Start the SRWS dewatering skid diaphragm pump.

13. Verify each LRWS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each LRWS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

30.02.XX System Level Tests 30.02.01.

This test is performed after the completion of Test 29.02.01 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

1. Verify LRWS can process a 1) Align LRWS to receive PZR 1) The LRWS degasifier removes gaseous waste stream. gaseous waste from the PZR condensable gases and vents during hot functional testing. waste to the RBVS or GRWS.

2) Process the PZR gaseous waste 2) The LRWS degasifier liquid through the LRWS degasifier. transfer pumps transfer the liquid

3) Purge the degasifier with nitrogen condensate waste to the low following operation. conductivity waste collection tanks.

3) LRWS degasifier is purged with nitrogen.

NuScale US460 SDAA 14.2-75 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-30: Test # 30 Liquid Radioactive Waste System (Continued) 30.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify LRWS can process a liquid Align LRWS to receive liquid waste 1) The waste treatment streams are waste stream. from a liquid waste stream. successfully processed through

1) Process the liquid waste stream the following processes:

through the LCW waste process.

filtration

2) Process the liquid waste stream

tubular filtration skid through the HCW process.

LCW processing skid

HCW processing skid

drum dryer skid

demineralization

transfer to LCW or HCW sample tanks

transfer from LCW or HCW sample tanks to the utility water system discharge basin.

NuScale US460 SDAA 14.2-76 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-31: Test # 31 Gaseous Radioactive Waste System Preoperational test is required to be performed once.

The GRWS is described in Section 11.3 and the functions verified by this test or another preoperational test are:

System Function System Function Categorization Function Verified by Test #

1. The GRWS supports the LRWS nonsafety-related 31.02.01 by receiving and collecting potentially radioactive and hydrogen-bearing waste gases that require processing before release to the environment.

2. The GRWS supports the CES by nonsafety-related 31.02.01 receiving and collecting potentially 36.02.02 radioactive and hydrogen-bearing waste gases that require processing before release to the environment.

3. The NDS supports the GRWS by nonsafety-related 31.02.01 providing nitrogen for purging of Table 14.2-12 component-level tests the GRWS.

31.00.XX Prerequisites

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

31.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each GRWS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each GRWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each GRWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify GRWS valves automatically 1) Initiate a real or simulated high MCR display and local, visual operate to maintain vessel GRWS moisture separator level. observation indicate the following:

volume. 2) Initiate a real or simulated low 1) The moisture separator drain GRWS moisture separator level. valve is open.

2) The moisture separator drain valve is closed.

05. Verify GRWS inlet isolation valves 1) Simulate a GRWS inlet stream MCR display and local, visual automatically close and nitrogen oxygen concentration high signal. observation indicate the following:

purge valve opens on high inlet 1) The inlet stream isolation valves stream oxygen concentration. are closed.

2) The nitrogen purge valve is open.

06. Verify GRWS isolates upon loss of 1) Simulate a loss of RWBVS exhaust 1) MCR display and local, visual RWBVS exhaust flow. flow. observation indicate the GRWS isolation valves are closed.

NuScale US460 SDAA 14.2-77 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-31: Test # 31 Gaseous Radioactive Waste System (Continued)

07. Verify radiation isolation of GRWS 1) Initiate a real or simulated GRWS MCR display and local, visual charcoal decay beds upon train A decay bed discharge flow observation indicate the following:

detection of decay bed discharge high radiation signal. 1) The following GRWS valves are flow high radiation level. 2) Initiate a real or simulated GRWS closed:

train B decay bed discharge flow a. GRWS charcoal decay bed skid A high radiation signal. outlet isolation valve

b. GRWS charcoal decay bed skid A inlet isolation valve

c. GRWS to Radioactive Waste Building HVAC system (RWBVS) exhaust upstream isolation valve

d. GRWS to RWBVS exhaust downstream isolation valve

2) The following GRWS valves are closed:

a. GRWS charcoal decay bed skid B outlet isolation valve

b. GRWS charcoal decay bed skid B inlet isolation valve

c. GRWS to Radioactive Waste Building HVAC system (RWBVS) exhaust upstream isolation valve

d. GRWS to RWBVS exhaust downstream isolation valve

[ITAAC 03.09.04]

(items 1 and 2)

08. Verify radiation isolation of GRWS 1) Initiate a real or simulated GRWS 1) MCR display and local, visual discharge to the RBVS exhaust discharge to the RBVS exhaust observation indicate the GRWS upon detection of a high radiation high radiation signal. discharge to the RBVS exhaust level. isolation valves are closed.

[ITAAC 03.09.04]

09. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from a GRWS grab flow through the grab sampling obtained.

sample device indicated on the device.

GRWS piping and instrumentation diagram.

10. Verify each GRWS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each GRWS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

31.02.XX System Level Test 31.02.01.Test Test Objective Test Method Acceptance Criteria

1. Verify GRWS can process a 1) Align GRWS to receive gaseous 1) The gaseous waste stream is gaseous waste stream and waste from a gaseous waste successfully processed through nitrogen stream. stream. Process the gaseous the following processes:

waste stream through the gaseous

gas cooler waste process.

moisture separator

2) Align GRWS charcoal drying

charcoal guard bed heater to receive nitrogen from

charcoal decay beds NDS. Process nitrogen through the

RWB exhaust charcoal drying process. 2) Nitrogen is successfully processed through the charcoal drying heater.

NuScale US460 SDAA 14.2-78 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-32: Test # 32 Solid Radioactive Waste System Preoperational test is required to be performed once.

The SRWS is described in Section 11.4 and the functions verified by this test or another preoperational test are:

System Function System Function Categorization Function Verified by Test #

1. The SRWS supports the LRWS by nonsafety-related 32.02.01 receiving spent resin and carbon 32.02.04 bed from LRWS processing skids. 32.02.06

2. The SRWS supports the CVCS by nonsafety-related 32.02.02 receiving spent resin from CVCS 32.02.05 ion exchange vessels.

3. The SRWS supports the PCWS nonsafety-related 32.02.03 by receiving spent resin and 32.02.05 sludge from PCWS ion exchange vessels.

4. The SRWS supports the CRVS by nonsafety-related 32.02.08 receiving exhausted HEPA filters to be compacted and shipped off site.

5. The SRWS supports the RWBVS nonsafety-related 32.02.08 by receiving exhausted HEPA filters to be compacted and shipped off site.

6. The SRWS supports the RBVS by nonsafety-related 32.02.08 receiving exhausted HEPA filters and charcoal bed from RBVS and CRVS, to be compacted and shipped off site.

7. The SRWS supports the GRWS nonsafety-related 32.02.08 by receiving contaminated or exhausted charcoal beds, packaging the waste in approved containers and shipping it to a licensed facility.

8. The LRWS supports the SRWS by nonsafety-related 32.02.07 receiving and processing liquid 30.02.02 radioactive waste from the SRWS dewatering skid.

NuScale US460 SDAA 14.2-79 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-32: Test # 32 Solid Radioactive Waste System (Continued) 32.00.XX Prerequisites

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

32.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each SRWS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each SRWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each SRWS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify each SRWS pump can be Align the SRWS to allow for pump 1) MCR display and local, visual started and stopped remotely. operation. observation indicate each pump

1) Start and stop each pump from the starts and stops. Audible and MCR. visible water hammer are not observed when the pump starts.

05. Verify the speed of each SRWS Align the SRWS to provide a flow path 1) MCR display indicates the speed variable-speed pump can be to operate a selected pump. of each obtains both minimum and manually controlled. 1) Vary the SRWS pump speed from maximum pump speeds.

minimum to maximum from the MCR.

06. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from an SRWS grab flow through the grab sampling obtained.

sample device indicated on the device.

SRWS piping and instrumentation diagram.

07. Verify each SRWS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each SRWS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

32.02.XX System Level Tests 32.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify spent resin from the LRWS Align the LRWS and SRWSs to 1) The waste management control demineralizers can be transferred transfer LRWS demineralizer resin to room (WMCR) displays and local, to the SRWS phase separator an SRWS phase separator tank. visual observation verifies LRWS tanks. 1) Start a phase separator transfer demineralizer resins transferred to pump. an SRWS phase separator tank.

32.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify spent resin from the CVCS Align the CVCS and SRWSs to 1) WMCR displays and local, visual ion exchangers can be transferred transfer CVCS ion exchanger resin to observation verifies CVCS ion to the SRWS spent resin storage an SRWS spent resin storage tank. exchanger resin transferred to an tanks. 1) Start an SRWS spent resin storage SRWS spent resin storage tank.

tank transfer pump.

NuScale US460 SDAA 14.2-80 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-32: Test # 32 Solid Radioactive Waste System (Continued) 32.02.03.

Test Objective Test Method Acceptance Criteria

1. Verify spent resin from the PCWS Align the PCWS and SRWSs to 1) WMCR displays and local, visual demineralizers can be transferred transfer PCWS demineralizer resin to observation verifies PCWS to the SRWS spent resin storage an SRWS spent resin storage tank. demineralizer resins transferred to tanks. 1) Start an SRWS spent resin storage an SRWS spent resin storage tank transfer pump. tank.

32.02.04.

Test Objective Test Method Acceptance Criteria

1. Verify spent resin from the SRWS Align an SRWS phase separator tank 1) WMCR displays and local, visual phase separator tanks can be and the SRWS dewatering station to observation verifies phase transferred to a dewatering station transfer spent resin to the dewatering separator tank resins are high integrity container (HIC). station HIC using the SAS. transferred to a dewatering station

1) Open SAS isolation valve to the HIC.

SRWS phase separator tank.

32.02.05.

Test Objective Test Method Acceptance Criteria

1. Verify spent resin from the SRWS Align an SRWS spent resin storage 1) WMCR displays and local, visual spent resin storage tanks can be tank and the SRWS dewatering station observation verifies spent resin transferred to a dewatering station to transfer spent resin to the storage tank resins are transferred HIC. dewatering station HIC using SAS air. to a dewatering station HIC.

1) Open SAS isolation valve to the spent resin storage tank.

32.02.06.

Test Objective Test Method Acceptance Criteria

1. Verify granulated activated 1) Align a LRWS and SRWS to GAC 1) WMCR displays and local, visual charcoal (GAC) from the LRWS to the dewatering station HIC using observation verifies spent resin granulated activated charcoal filter the clean in place system. storage tank resins are transferred can be transferred to a dewatering to a dewatering station HIC.

station HIC.

32.02.07.

Test Objective Test Method Acceptance Criteria

1. Verify the dewatering skid pump 1) Align the dewatering skid pump to 1) Free-standing water in the HIC removes standing water in the HIC an LRWS high conductivity waste has been removed.

with spent resin in the dewatering tank and start the dewatering skid station HIC. pump.

32.02.08.

Test Objective Test Method Acceptance Criteria

1. Verify the SRWS waste 1) Place solid radioactive waste in 1) The waste has been compacted.

compactor compacts solid compactor and start compactor.

radioactive waste.

NuScale US460 SDAA 14.2-81 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-33: Test # 33 Chemical and Volume Control System Preoperational test is required to be performed for each NPM.

The CVCS is described in Section 9.3.4 and the functions verified by this test, other preoperational tests, and power ascension testing are:

System Function System Function Categorization Function Verified by Test #

1. The CVCS supports the RCS by nonsafety-related 33.02.01 providing primary coolant makeup. 94.03.01

2. The CVCS supports the RCS by nonsafety-related 33.02.01 providing primary coolant letdown. 94.03.01

3. The CVCS supports the RCS by nonsafety-related 33.02.02 providing PZR spray flow for RCS 94.03.01 pressure control.

4. The CVCS supports the RCS by nonsafety-related 33.02.03 changing the boron concentration of the primary coolant.

5. The BAS supports the CVCS by nonsafety-related 33.02.03 providing uniformly mixed borated water on demand.

6. The LRWS supports the CVCS by nonsafety-related 33.02.01 receiving and processing primary 30.02.02 coolant from CVCS letdown.

The CVCS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

7. The CVCS supports the ECCS nonsafety-related 29.02.01 valves by providing water to reset the ECCS valves.

8. The CVCS supports the RCS by nonsafety-related 29.02.01 heating primary coolant.

9. The CVCS supports the RCS by safety-related 56.02.04 isolating dilution sources.

10.The CVCS supports the RCS by nonsafety-related 56.02.08 providing primary coolant makeup in beyond design-basis events.

33.00.XX Prerequisites

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

02. Verify a pump curve test is completed and approved for the CVCS pumps.

03. Component Level Tests 33.01.04, 33.01.05 and 33.01.06 must be performed under preoperational test conditions that approximate design-basis temperature, differential pressure, and flow conditions to the extent practicable, consistent with preoperational test limitations.

33.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each CVCS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each CVCS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

03. Verify each CVCS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

NuScale US460 SDAA 14.2-82 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-33: Test # 33 Chemical and Volume Control System (Continued)

04. Verify each CVCS American 1) Operate each valve from the MCR. 1) MCR display verifies the valve Society of Mechanical Engineers opens and closes under (ASME) Code Class 3 preoperational temperature, air-operated valve changes differential pressure, and flow position under preoperational conditions.

temperature, differential pressure, [ITAAC 02.02.01]

and flow conditions.

05. Verify each CVCS ASME Code Place each valve in its non-safe 1) MCR display and local, visual Class 3 air-operated valve fails to position. observation indicate each valve its safe position on loss of air 1) Isolate and vent air to the valve. fails to its safe position under under preoperational temperature, preoperational temperature, differential pressure, and flow differential pressure, and flow conditions. conditions.

[ITAAC 02.02.02]

06. Verify each CVCS ASME Code Place each valve in its non-safe 1) MCR display and local, visual Class 3 air-operated valve fails to position. observation indicate each valve its safe position on loss of 1) Isolate electrical power to the fails to its safe position under electrical power to its solenoid valve. preoperational temperature, under preoperational temperature, differential pressure, and flow differential pressure, and flow conditions.

conditions. [ITAAC 02.02.02]

07. Verify each CVCS pump can be Align the CVCS to allow for pump 1) MCR display and local, visual started and stopped remotely. operation. observation indicate the pump

1) Start and stop each CVCS pump starts and stops. Audible and from the MCR. visible water hammer are not observed when the pump starts.

08. Verify the speed of each CVCS Align the CVCS to provide a flow path 1) MCR display indicates the speed variable-speed pump can be to operate a selected pump. of each obtains both minimum and manually controlled. 1) Vary the CVCS pump speed from maximum pump speeds.

minimum to maximum from the MCR.

09. Verify each CVCS operating Align the CVCS to allow for pump 1) MCR display and local, visual makeup pump automatically stops operation. Place a makeup pump in observation indicate the operating to protect the pump and the service. pump stops and the standby pump standby pump starts. 1) Initiate a simulated makeup pump starts. Audible and visible water trip. hammer are not observed when the pump starts.

10. Verify each CVCS recirculation Align the CVCS to allow for pump 1) MCR display and local, visual pump automatically stops to operation. Place a recirculation pump observation indicate the operating protect the pump and the standby in service. pump stops and the standby pump pump starts. 1) Initiate a simulated recirculation starts.

pump trip.

11. Verify CVCS letdown flow isolates 1) Initiate a simulated CVCS high 1) MCR display and local, visual on high flow to protect plant letdown flow signal. observation indicate the LRWS equipment. letdown flow control valve and LRWS letdown isolation valves (3) are closed.

NuScale US460 SDAA 14.2-83 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-33: Test # 33 Chemical and Volume Control System (Continued)

12. Verify ion exchanger isolation on 1) Initiate a simulated high 1) MCR display and local, visual non-regenerative heat exchanger non-regenerative heat exchanger observation indicate the following:

high outlet temperature to protect outlet temperature signal. a. CVCS purification bypass diverting plant equipment. valve is in the bypass position.

b. Mixed bed ion exchanger A inlet isolation valves (2) are closed.

c. Auxiliary ion exchanger inlet isolation valve is closed.

d. Cation exchanger inlet isolation valve is closed.

13. Verify the CVCS automatically 1) Initiate a real or simulated high MCR display verifies the following:

responds to mitigate a release of radiation signal for the RCS 1) CVCS RCS discharge to process radioactivity. discharge flow to the regenerative sampling isolation valve closed.

heat exchanger. [ITAAC 02.07.02]

14. Verify each CVCS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each CVCS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

33.02.XX System Level Tests 33.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify proper operation of the This test is performed in conjunction 1) MCS data indicate that automatic automatic PZR level control. with Turbine Generator System Test PZR letdown maintained PZR 29.02.01, which heats the RCS from level at setpoint as described in ambient conditions to no less than Section 9.3.4.

345°F but as high as reasonably 2) MCS data indicate that the PZR achievable. level control results in CVCS

1) Place PZR level control in makeup to the RCS to increase automatic operation during RCS PZR level to the target setpoint as heatup to demonstrate automatic described in Section 9.3.4.

letdown. Use the MCS data historian to review PZR level at maximum-obtained RCS temperature.

2) To raise PZR level, use MCS automation and operator permission to increase to a target PZR level.

Note: PZR letdown level control is automatic; however, PZR makeup level control is automatic with consent of operator.

NuScale US460 SDAA 14.2-84 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-33: Test # 33 Chemical and Volume Control System (Continued) 33.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify proper operation of the This test is performed in conjunction 1) MCS data indicate automatic PZR automatic PZR pressure control. with Turbine Generator System Test heater operation raised PZR

2. Verify no dynamic effects caused 29.02.01, which heats the RCS from pressure to the setpoint as by changes in fluid flow. ambient conditions to no less than described in Section 9.3.4.

345°F but as high as reasonably 2) MCS data indicate automatic PZR achievable. spray valve operation lowered

1) Place PZR pressure control in PZR pressure to the spray valve automatic and raise pressure closure setpoint as described in setpoint to the normal operating Section 9.3.4.

band. 3) Water hammer indications:

2) Raise PZR pressure to the PZR a. Audible indications of water spray valve open setpoint. Use the hammer are not observed.

MCS data historian to review PZR b. No damage to pipe supports or pressure at maximum-obtained restraints.

RCS temperature. c. No damage to equipment.

d. No equipment leakage 33.02.03.

Test Objective Test Method Acceptance Criteria

1. Verify proper operation of CVCS This test is performed in conjunction 1) BAS storage tank sample boron automatic dilution and boration with Turbine Generator System Test concentration is within control. 29.02.01, which heats the RCS from specifications (as described in ambient conditions to no less than Section 9.3.4).

345°F but as high as reasonably 2) MCS data indicate that the dilution achievable. of the RCS results in a decreased Ensure that RCS low flow rate alarm is boron concentration within clear to ensure adequate mixing for acceptable limits of the target dilution and boration. concentration as described in

1) Place the BAS storage tank on Section 9.3.4.

recirculation and sample boron 3) MCS data indicate that the concentration. boration of the RCS results in a

2) Use the MCS automation and increased boron concentration operator permission to decrease to within acceptable limits of the a target RCS boron concentration. target concentration as described

3) Use the MCS and operator. in Section 9.3.4.

Permission to increase to a target RCS boron concentration.

NuScale US460 SDAA 14.2-85 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-34: Test # 34 Boron Addition System Preoperational test is required to be performed for each NPM.

The 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 #

1. The BAS supports the PCWS by nonsafety-related 34.02.01 providing borated water to the RXB pools.

The BAS function verified by other test is:

System Function System Function Categorization Function Verified by Test #

2. The BAS supports the CVCS by nonsafety-related 33.02.03 providing uniformly mixed borated water on demand.

34.00.XX Prerequisites

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

02. Verify a pump curve test is completed and approved for the BAS pumps.

34.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each BAS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each BAS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each BAS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify the BAS transfer pump can Align the BAS to allow for pump 1) MCR display and local, visual be started and stopped remotely. operation. observation indicate the pump

1) Start and stop the transfer pump starts and stops. Audible and from the MCR. visible water hammer are not observed when the pump starts.

05. Verify the BAS supply pump can Align the BAS to allow for pump 1) MCR display and local, visual be started and stopped remotely. operation. observation indicate the pump

1) Start and stop the supply pump starts and stops. Audible and from the MCR. visible water hammer are not observed when the pump starts.

06. Verify the speed of the BAS Align the BAS to provide a flow path to 1) MCR display indicates the speed variable-speed pumps can be operate a selected pump. of each pump obtains both manually controlled. 1) Vary the BAS pump speed from minimum and maximum pump minimum to maximum from the speeds. Audible and visible water MCR. hammer are not observed when the pump starts.

07. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from a BAS grab sample flow through the grab sampling obtained.

device. device.

NuScale US460 SDAA 14.2-86 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-34: Test # 34 Boron Addition System (Continued)

08. Verify each BAS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each BAS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

34.02.XX System Level Test 34.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the BAS automatically adds 1) Verify the BAS batch tank contains MCR displays and local, visual a specified quantity of borated a sufficient volume of water to observation verifies the following:

water from the BAS batch tank to conduct this test. 1) The BAS to PCWS valve initially the RXB pools. 2) Align the BAS and the PCWS to opens to supply water from the supply water from the BAS to the BAS to the PCWS.

PCWS pump suction. 2) The BAS to PCWS valve

3) Enter a BAS batch tank target level automatically closes when the to terminate batch operation to the BAS batch tank obtains the target spent fuel pool. level.

NuScale US460 SDAA 14.2-87 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-35: Test # 35 Module Heatup System Preoperational test is required to be performed for each NPM.

The MHS is described in Section 9.3.4. MHS functions are not verified by MHS tests. MHS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. The MHS supports the CVCS by nonsafety-related 29.02.01 adding heat to primary coolant.

35.00.XX Prerequisites

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

35.01.XX Component Level Tests

01. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is obtained.

obtained from an MHS grab flow through the grab sampling sample device. device.

02. Verify each MHS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each MHS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

35.02.XX System Level Tests None NuScale US460 SDAA 14.2-88 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-36: Test # 36 Containment Evacuation System Preoperational test is required to be performed for each NPM.

The CES is described in Sections 9.3.6 and 5.2.5 and the functions verified by this test or another preoperational test are:

System Function System Function Categorization Function Verified by Test #

1. The CES supports the CNTS by nonsafety-related 36.02.01 removing water vapor from the 36.02.02 containment vessel (CNV). 36.02.03

2. The CES supports the CNTS by nonsafety-related 36.02.01 condensing water vapor removed 36.02.02 from the CNV in the CES 36.02.03 condenser.

3. The CES supports the CNTS by nonsafety-related 36.02.01 removing non-condensable gases 36.02.02 from the CNV.

4. The CES supports the RCS by nonsafety-related 36.02.03 providing RCS leak detection monitoring capability.

5. The GRWS supports the CES by nonsafety-related 36.02.02 receiving and collecting potentially 31.02.01 radioactive and hydrogen-bearing waste gases that require processing before release to the environment.

36.00.XX Prerequisites

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

36.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each CES 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each CES air-operated Place each CES valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each CES air-operated Place each CES valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. CES air-operated valve.

04. Verify each CES pump can be 1) Start and stop each pump from the 1) MCR display and local, visual started and stopped remotely. MCR. observation indicate each pump starts and stops.

05. Verify the speed of each CES 1) Vary the speed of each pump from 1) MCR display indicates pump variable-speed pump can be the MCR and local control panel (if speed varies from minimum to manually controlled. design has local pump control). maximum pump speed.

06. Verify each CES pump Place a pump in operation. 1) MCR displays and local, visual automatically stops to protect 1) Initiate a real or simulated signal for observation verifies the pump plant equipment. each pump trip condition. stops.

07. Verify each CES pump suction Open the pump suction and discharge 1) Each pump suction and discharge and discharge valve automatically valves. valve closes on each real or closes to protect the CES 1) Initiate a real or simulated signal for simulated valve close condition.

equipment. each valve close conditions.

NuScale US460 SDAA 14.2-89 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-36: Test # 36 Containment Evacuation System (Continued)

08. Verify a local grab sample can be 1) Place the system in service to allow 1) A local grab sample is successfully obtained from a CES grab sample flow through the grab sampling obtained.

device indicated on the CES device.

piping and instrumentation diagram.

09. Verify each CES instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each CES on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

36.02.XX System Level Tests 36.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the automatic operation of 1) After the CFDS completes 1) The automated control establishes the CES to establish and maintain draindown of the CNV and the and maintains vacuum in the CNV design vacuum for the CNV. NPM is in hot functional testing, within design limit per place the CES in automatic Section 6.2.2.

operation.

36.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify radiation isolation and flow The NPM is in hot functional testing 1) diversion on high radiation level in with the RCS at normal operating a. The CES effluent flow path to the the CES. pressure. The CES is operating in RBVS is isolated and diverted to automatic control with a CNV GRWS.

steady-state vacuum pressure b. The CES effluent to process indicating the noncondensable gases sample panel isolation valve is have been removed from the CNV. closed.

1) Initiate a real or simulated high c. The CES purge air solenoid valves radiation signal for the CES to the vacuum pumps are closed.

vacuum pump discharge. [ITAAC 02.07.01]

2) Initiate a simulated high area (Items 1a through 1c) radiation signal for the GRWS d. The automated control maintains charcoal bed cubicle to the vacuum in the CNV.

individual valve being tested. 2) The CES to GRWS vapor condenser isolation valve is closed.

[ITAAC 02.07.01]

NuScale US460 SDAA 14.2-90 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-36: Test # 36 Containment Evacuation System (Continued) 36.02.03.

Test Objective Test Method Acceptance Criteria

1. Verify the CES level 1) The NPM is in hot functional testing 1) The CES detects a level increase instrumentation supports RCS with the RCS at normal operating in the CES sample tank, which leakage detection. pressure and the maximum correlates to a detection of an

2. Verify the CES pressure operating temperature achievable unidentified RCS leakage rate of instrumentation supports RCS by heating the RCS with the MHS. one gpm within one hour, by leakage detection. 2) The CES is operating in automatic providing an alarm signal to the control with a CNV steady-state MCR within one hour of the start of vacuum pressure indicating the water injection into the CNV noncondensable gasses have indicating the baseline leakage been removed from the CNV. rate has been exceeded.

3) Record the MCS baseline leakage [ITAAC 02.03.01]

rate into the CNV. 2) The CES detects a pressure

4) Isolate the CFDS to CNTS spool increase in the inlet pressure piece to allow test equipment to be instrumentation that correlates to a connected to the spool piece. detection of an unidentified RCS

5) Inject water at a flow rate less than leakage rate of one gpm within or equal to one gpm. one hour, by providing an alarm This test may be done in conjunction signal to the MCR within one hour with Test 36.02.02. of the start of water injection into the CNV indicating the baseline leakage rate has been exceeded.

[ITAAC 02.03.02]

NuScale US460 SDAA 14.2-91 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-37: Test # 37 Containment Flooding and Drain Preoperational test component level testing is required to be performed once. System level testing is required to be performed as indicated for each system level test.

The CFDS is described in Section 9.3.7 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 37.02.02 flooding the CNV in preparation for refueling operations.

2. The CFDS supports the CNTS by nonsafety-related 37.02.01 draining the CNV in preparation for startup operations.

The CFDS function verified by another test is:

3. The CFDS supports the RCS by nonsafety-related 56.02.08 providing borated coolant inventory for the removal of core heat during a beyond-design-basis accident.

37.00.XX Prerequisites

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

37.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each CFDS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each CFDS air-operated Place each CFDS valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

03. Verify each CFDS air-operated Place each CFDS valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. CFDS air-operated valve.

04. Verify each CFDS pump can be 1) Start and stop each pump from the 1) MCR display and local, visual started and stopped remotely. MCR. observation indicate each pump starts and stops. Audible and visible water hammer are not observed when the pump starts.

05. Verify each CFDS pump Place a pump in operation. 1) MCR displays and local, visual automatically stops to protect 1) Initiate a real or simulated signal for observation verifies the pump plant equipment. each pump trip condition. stops.

06. Verify each CFDS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each CFDS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

NuScale US460 SDAA 14.2-92 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-37: Test # 37 Containment Flooding and Drain (Continued) 37.02.XX System Level Tests 37.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the CFDS can automatically 1) Drain the CNTS using CFDS 1) The CNTS is drained using CFDS drain the CNTS. automatic operation and designed automatic controls.

manual operation. (This test is required to be performed for each NPM.)

37.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify the CFDS can automatically 1) Flood the CNTS using CFDS 1) The CNTS is flooded using CFDS flood the CNTS. automatic operation and designed automatic controls.

manual operation. (This test is required to be performed for each NPM.)

37.02.03.

Test Objective Test Method Acceptance Criteria

1. Verify the CFDS automatically 1) While the CFDS is draining the 1) The CFDS containment drain responds to mitigate a release of CNTS, initiate a real or simulated separator gaseous discharge to radioactivity. high radiation signal on the RBVS isolation valve is closed.

gaseous effluent of the CFDS [ITAAC 03.09.05]

containment drain separator tank.

NuScale US460 SDAA 14.2-93 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-38: Test # 38 Containment System 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 38.02.01 providing a barrier to contain 38.02.02 mass, energy, and fission product release from a degradation of the reactor coolant pressure boundary.

2. The CNTS supports the ECCS safety-related 38.02.01 operations by providing a sealed containment.

3. The ECCS supports CNTS by safety-related 38.02.01 providing a portion of the containment boundary for maintaining containment integrity.

The CNTS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

4. The CNTS supports the DHRS by safety-related 56.02.04 closing CIVs for the MSS and FWS when actuated by the MPS for DHRS operation.

5. The CNTS supports the RCS by safety-related 56.02.04 closing the CIVs for PZR spray, RCS injection, RCS discharge, and reactor pressure vessel (RPV) high point degasification when actuated by the MPS for RCS isolation.

6. The CNTS supports the RXB by safety-related 56.02.04 providing a barrier to contain mass, energy, and fission product release by closure of the CIVs upon a containment isolation signal.

7. The CNTS supports the Reactor nonsafety-related, risk-significant 45.02.01 Building crane (RBC) by providing 45.02.02 lifting attachment points that the RBC can connect to so that the module can be lifted.

8. The CNTS supports the MPS by nonsafety-related 59.02.02 providing PAM nonsafety-related information signals.

NuScale US460 SDAA 14.2-94 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-38: Test # 38 Containment System (Continued) 38.00.XX Prerequisites

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

38.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each hydraulic skid 1) Verify each hydraulic skid supplies 1) Pump maintains required system supplies sufficient pressure for sufficient pressure for valve pressure.

valve operation. operation.

02. Verify each CNTS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each CNTS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

38.02.XX System Level Tests 38.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the leak tightness of the 1) Perform 10 CFR Part 50, Appendix 1) Local leak rate tests are CNTS. J local leak rate tests (Type B and completed on containment Type C tests) of the CNTS in penetrations listed in Table 6.2-4 accordance with the guidance that require Appendix J, Type B or provided in ANSI/ANS 56.8, RG C testing.

1.163, and NEI 94-01. [ITAAC 02.01.07]

38.02.02.

Test 38.02.02 is performed at hot functional conditions.

Test Objective Test Method Acceptance Criteria

1. Verify the CNTS safety-related 1) The check valves are tested in 1) Each CNTS safety-related check check valves change position accordance with the requirements valve strokes fully open and under design temperature, of ASME OM code, ISTC-5220, closed under forward and reverse differential pressure, and flow. check valves. flow conditions, respectively.

[ITAAC 02.01.21]

NuScale US460 SDAA 14.2-95 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-39: Test # 39 Reactor Coolant System Preoperational test is required to be performed for each NPM.

The RCS is described in Section 5.4 and the RCS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. The RCS supports the MPS by safety-related, risk-significant 56.02.01 providing instrument information signals for MPS actuation.

2. The RCS supports the MPS by safety-related 56.02.01 providing instrument information signals for low temperature overpressure protection (LTOP) actuation.

3. The RCS supports the MPS by nonsafety-related 59.02.02 providing PAM instrument information signals.

39.00.XX Prerequisites

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

39.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each RCS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each RCS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

39.02.XX System Level Tests None NuScale US460 SDAA 14.2-96 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-40: Test # 40 Emergency Core Cooling System Preoperational test is required to be performed for each NPM.

System Level Test 40.02.01 is only required to be performed once for the first NPM tested. This test supports FOAK testing as described in Section 14.2.3.3.

The ECCS is described in Section 6.3, and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The ECCS supports the RCS by safety-related 40.02.01 opening the ECCS reactor vent 56.02.02 valves (RVVs) and reactor recirculation valves (RRVs) when their respective trip valve is actuated by the MPS.

2. The ECCS supports the RCS by safety-related 40.02.01 providing recirculated coolant from 56.02.02 the containment to the RPV for the removal of core heat.

The ECCS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

3. The ECCS supports the RCS by safety-related 56.02.02 providing LTOP for maintaining the reactor coolant pressure boundary.

4. The ECCS supports the CNTS by safety-related 38.02.01 providing a portion of the containment boundary for maintaining containment integrity.

5. The ECCS supports MPS by nonsafety related 59.02.01 providing PAM instrument information signals.

40.00.XX Prerequisites

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

40.01.XX Component Level Tests None NuScale US460 SDAA 14.2-97 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-40: Test # 40 Emergency Core Cooling System (Continued) 40.02.XX System Level Test 40.02.01.

Test 40.02.01 is performed at hot functional testing to allow ECCS actuation at elevated RCS pressure and temperature conditions, starting just above the inadvertent actuation block (IAB).

The RCS is heated to the highest temperature achievable by MHS heating. These hot functional testing conditions provide the highest temperature conditions that can be achieved before fuel load. The RCS level is within the expected range of module operation, near the low end of the normal operating range for hot zero power (HZP) conditions. This test can be performed concurrently with Test 56.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify ECCS RRC valves open 1) Ensure RCS pressure is as close 1) ECCS RRVs open below IAB below high RCS pressure IAB to, but above, the IAB RCS setpoint.

setpoint. pressure threshold as practicable. 2) RPV riser level remains above the

2. Verify the RPV liquid level remains 2) Ensure RCS temperature is at the top of the core.

above the top of the core during maximum temperature achievable 3) CNV pressure remains within and following ECCS actuation. by heating the RCS using MHS upper and lower bounds

3. Verify the heat removal capacity of heating. calculated using safety analysis the ECCS, operating with the 3) Ensure RCS level is as low in the methods, while accounting for test CNV, is consistent with the design normal operating band as is initial conditions and basis. practically achievable for the instrumentation uncertainty.

established plant conditions. [ITAAC 02.01.14]

4) Manually initiate ECCS from the [ITAAC 02.01.19]

MCR.

5) Allow RPV riser level and CNV level to become relatively stable.

NuScale US460 SDAA 14.2-98 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-41: Test # 41 Decay Heat Removal System Preoperational test is required to be performed for each NPM. System Test 41.02.01 is required to be performed once for the first NPM tested. This test supports FOAK testing described in Section 14.2.3.3.

The DHRS is described in Section 5.4.3. FOAK 41.02.01 is described in Section 5.4.3. 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 56.02.04 opening the DHRS actuation 98.03.01 valves for DHRS operation.

2. The DHRS supports the MPS by safety-related 56.02.01 providing MPS actuation instrument information signals.

3. The DHRS supports the MPS by nonsafety-related 59.02.02 providing PAM instrument information signals.

4. The UHS supports the DHRS by safety-related 98.03.01 accepting the heat from the DHRS heat exchanger.

41.00.XX Prerequisites

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

41.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each DHRS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each DHRS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

41.02.XX System Level Test 41.02.01.

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

1. Verify DHRS removes heat from 1) Verify RCS is at normal operating 1) DHRS cooldown of RCS meets the RCS. pressure and the RCS has design-basis requirements.

achieved the maximum temperature achievable by warming the RCS using MHS heating.

2) Open DHRS actuation valves and close containment isolation valves by initiating a containment isolation via MPS.

3) Allow the RCS to cool down less than 345 degrees.

4) Compare RCS cooldown rate to test analysis conducted using the code of record as described in Section 5.4.3.

NuScale US460 SDAA 14.2-99 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-42: Test # 42 In-Core Instrumentation System Preoperational test is required to be performed for each NPM.

The in-core instrumentation system (ICIS) is described in Section 7.0.4 and the function verified by this test and power ascension testing is:

System Function System Function Categorization Function Verified by Test #

1. The ICIS supports the MPS by nonsafety-related 42.02.01 providing reactor core system 87.03.01 (RXCS) temperature information.

The ICIS functions verified by another test is:

System Function System Function Categorization Function Verified by Test #

2. The ICIS supports the MPS by nonsafety-related 59.02.02 providing PAM instrument information signals.

42.00.XX Prerequisites

01. The ICIS instrument strings are inserted into the core.

02. Verify an instrument calibration is performed on all ICIS thermocouples by cross-calibrating the thermocouple to the RCS narrow range resistance temperature detectors (RTDs) before RCS heatup.

42.01.XX Component Level Tests None 42.02.XX System Level Test 42.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify proper temperature 1) Heat the RCS from ambient 1) MCS data indicate that the ICIS indication is obtained from the conditions to the maximum RCS thermocouples respond properly.

ICIS thermocouples. temperature that can be obtained by the MHS.

2) Use the MCS data historian to cross-check the ICIS thermocouples to each other and the RCS narrow-range and wide range RTDs.

NuScale US460 SDAA 14.2-100 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-43: Test # 43 Module Assembly Equipment Preoperational test is required to be performed once.

The module assembly equipment (MAE) consists of module import trolley, the upender, and the inspection rack.

System Function System Function Categorization Function Verified by Test #

1. MAE supports the NPM actively nonsafety-related component-level tests by providing material handling to allow its transport in the horizontal orientation to travel from outside the RXB to its interior and to rotate it to operational orientation.

43.00.XX Prerequisites 01.An MAE factory acceptance test (FAT) is successfully completed and approved, if required.

43.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify the operation of MAE 1) Actuate or simulate actuation of the 1) The MAE equipment controls limit controls that limit motion and interlocks. motion and speed per design.

speed.

(This test may be performed as part of SAT.)

43.02.XX System-Level Tests None NuScale US460 SDAA 14.2-101 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-44: Test # 44 Fuel Handling Equipment Preoperational test is 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 nonsafety-related 44.02.01 fuel by providing ability to visually 44.02.02 inspect fuel.

2. The FHE system supports the nonsafety-related 44.02.03 RXCS by moving fuel within the 44.02.04 core.

3. The FHE system supports the nonsafety-related 44.02.04 spent fuel storage system by moving fuel into the spent fuel storage system.

44.00.XX Prerequisites 01.An FHE system FAT is successfully completed and approved.

02. A rated-load test is successfully completed and approved on the FHE system on the following equipment in accordance with ASME NOG-1 paragraph 7423.

a. Fuel handling machine (FHM) main hoist

b. FHM auxiliary hoists

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

44.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify the operation of FHE 1) Actuate or simulate actuation of the 1) The FHE equipment controls limit controls that limit motion and interlocks. motion and speed per design.

speed.

02. Verify each FHE instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each FHE on an MCS or PCS display, or is display if the FHE instrument is system transmitter. recorded by the applicable control designed to be displayed on an system historian.

MCR workstation.

(Test not required if the instrument calibration verified the MCS or PCS display.)

44.02.XX System Level Tests 44.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the proper operation of the 1) Transfer a dummy fuel assembly 1) A dummy fuel assembly is new fuel jib crane. from its receipt shipping container successfully transferred to the new to the new fuel inspection stand fuel inspection stand.

and from the new fuel inspection 2) A dummy fuel assembly is stand to the new fuel elevator. successfully transferred to the new fuel elevator.

44.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify the proper operation of the 1) Lower a dummy fuel assembly in 1) A dummy fuel assembly is new fuel elevator. the new fuel elevator. successfully lowered to the position where it can be retrieved by the FHM mast.

NuScale US460 SDAA 14.2-102 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-44: Test # 44 Fuel Handling Equipment (Continued) 44.02.03.

Test Objective Test Method Acceptance Criteria

1. Verify the proper operation of the 1) Transfer the dummy fuel assembly 1) The dummy fuel assembly is FHM. from the new fuel elevator to the successfully transferred to the FHM mast. FHM mast.

2) Transfer the dummy fuel assembly 2) The dummy fuel assembly is from the new fuel elevator location successfully transferred to its to a designated RXCS location. designated core location and

3) Seat the dummy fuel assembly. partially inserted.

3) The dummy fuel assembly is fully seated.

44.02.04.

Test Objective Test Method Acceptance Criteria

1. Verify the proper operation of the 1) Withdraw the dummy fuel 1) The dummy fuel assembly is FHM. assembly to a position where the successfully transferred to its FHM can automatically transfer the designated storage location and assembly. partially inserted.

2) Transfer the dummy fuel assembly 2) The dummy fuel assembly is fully from the RXCS to a designated seated.

spent fuel storage location.

(Manual operation of the fuel assembly is required for final fuel insertion.)

3) Seat the dummy fuel assembly.

44.02.05.

Test Objective Test Method Acceptance Criteria

1. Verify the FHM maintains at least 1) Perform a test of the FHM mast 1) The FHM maintains at least 10 10 feet of water above the top of mechanical stop limit switch. feet of water above the top of the the fuel assembly when lifted to its fuel assembly when lifted to its maximum height with the pool maximum height with the pool level at the lower limit of the level at the lower limit of the normal operating low water level. normal operating low water level.

[ITAAC 03.04.05]

44.02.06.

Test Objective Test Method Acceptance Criteria

1. The new fuel jib crane hook 1) Using the new fuel jib crane hook 1) The new fuel jib crane interlocks movement is limited to prevent attempt to transfer a dummy fuel prevent the crane from carrying a carrying a fuel assembly over the assembly or new fuel assembly fuel assembly over the spent fuel fuel storage racks in the spent fuel over the fuel storage racks in the racks.

pool. spent fuel pool. [ITAAC 03.04.06]

NuScale US460 SDAA 14.2-103 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-45: Test # 45 Reactor Building Cranes Preoperational test is required to be performed once unless otherwise noted in the test.

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 45.02.01 providing structural support and 45.02.02 mobility while moving from refueling, inspection and operating bay.

2. The MAE bolting supports the nonsafety-related 45.02.02 CNTS by providing material handling to allow for disassembly and reassembly of the CNV lower flange.

3. The MAE bolting supports the nonsafety-related 45.02.02 RPV 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, risk-significant 45.02.01 providing lifting attachment points 45.02.02 that the RBC can connect to so that the module can be lifted.

45.00.XX Prerequisites 01.An RBC site acceptance test is completed and approved.

02. A rated-load test is completed and approved on the RBC on the NPM top support structure in accordance with ASME NOG-1 paragraph 7423.

a. RBC main hoist

b. RBC auxiliary hoists

c. RBC wet hoist

03. A rated-load test is completed and approved on the NPM top support structure in accordance with ANSI N14.6.

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

45.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify RBC controls that limit RBC 1) Actuate or simulate actuation of the 1) Local visual observation indicates motion and speed. RBC interlocks. that the interlocks limit RBC motion and speed.

02. Verify RBC remains in current Initiate the following real or simulated 1) Local visual observation indicates position on loss of control or signals: that the bridge, trolley, main hoist, power or seismic event. 1) Loss of control. wet hoist, auxiliary hoist trolley and

2) Loss of power. auxiliary hoist brakes are set.

3) Seismic switch actuation.

03. Verify each RBC instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each RBC on an MCS or PCS display, or is display if the RBC instrument is system transmitter. recorded by the applicable control designed to be displayed on an system historian.

MCR workstation.

(Test not required if the instrument calibration verified the MCS or PCS display.)

NuScale US460 SDAA 14.2-104 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-45: Test # 45 Reactor Building Cranes (Continued) 45.02.XX System Level Tests 45.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify RBC load path and removal Place the lower block assembly on the 1) The bridge and trolley speeds do of an NPM from a reactor bay. RBC. Lift an NPM and move the RBC not exceed maximum design

2. Verify RBC load path and with the attached NPM to its design speeds.

installation of an NPM in a reactor home location. 2) The bridge and trolley does not bay. 1) Use the RBC semi-automatic move at the same time.

programmed controls to install the 3) The bridge and trolley maximum NPM in the lead NPM bay location allowable speed is toggled from and return the RBC to the design full-speed to microspeed when the home location. RBC hook gets within the design

2) Use the RBC semi-automatic distance of a predefined reference programmed controls to retrieve location.

the NPM from the lead NPM bay 4) The main hoist only moves within location and return the RBC with the predefined elevation zones.

attached module to the design 5) The NPM is positioned at the home location. design rotation at predefined Repeat this sequence for each NPM reference locations.

installation. 6) The NPM is fully seated in the reactor bay receiver. (Acceptance Criteria (1) through (4) only need to be satisfied for the first performance of the test.

Acceptance Criteria (5) and (6) need to be satisfied for each NPM.

NuScale US460 SDAA 14.2-105 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-45: Test # 45 Reactor Building Cranes (Continued) 45.02.02.

Test Objective Test Method Acceptance Criteria

1. The RBC is at the design home 1)

a. Verify the NPM can be location with an NPM attached to the a. The NPM is disassembled using disassembled using the CNV lower block assembly. the CNV support stand and the support stand and the RPV 1) Use the RBC semi-automatic RPV support stand and associated support stand and associated programmed controls to move the tooling.

tooling. NPM from the design home b. The RBC semi-automatic controls

b. Verify the RBC semi-automatic location to the CNV support stand are used to transport the NPM controls can be used to transport and seat the NPM lower CNV in the through the disassembly process.

the NPM through the disassembly CNV support stand. De-tension 2) process. and remove the lower CNV closure a. The NPM is assembled using the

2. bolts. CNV support stand and the RPV

a. Verify the NPM can be assembled Use the RBC semi-automatic support stand and associated using the CNV support stand and programmed controls to move the tooling.

the RPV support stand and NPM from the CNV support stand b. The RBC semi-automatic controls associated tooling. to the RPV support stand and seat are used to transport the NPM

b. Verify the RBC semi-automatic the NPM in the RPV support stand. through the assembly process.

controls can be used to transport De-tension and remove the lower the NPM through the assembly RPV closure bolts.

process. Use the RBC semi-automatic 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 lower block assembly from the upper NPM and move the RBC and lower block assembly from the module inspection rack to the design home location.

NuScale US460 SDAA 14.2-106 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-45: Test # 45 Reactor Building Cranes (Continued)

2) Use the RBC semi-automatic programmed controls to move the NPM and lower block assembly from the design home location to the module inspection rack and attach the upper NPM to the lower block assembly.

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.

NuScale US460 SDAA 14.2-107 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-46: Test # 46 Process Sampling System 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 46.02.01 normal operations by providing sampling and analysis of reactor coolant discharge (letdown) liquid.

2. The PSS supports the CVCS by nonsafety-related 46.02.01 providing sampling of reactor coolant at process points in the CVCS.

3. The PSS supports the CNTS nonsafety-related 46.02.02 during normal operations by providing sampling of containment gas and analysis of hydrogen and oxygen concentration in containment.

4. The PSS supports the FWS by nonsafety-related 46.02.03 providing sampling and analysis of condensate and FW.

5. The PSS supports the MSS by nonsafety-related 46.02.03 providing sampling and analysis of MS.

6. PSS supports the CNTS during nonsafety-related 46.02.02 accident condition by providing containment atmosphere monitoring and analysis of hydrogen and oxygen concentration to respond to emergencies.

46.00.XX Prerequisites

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

46.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each PSS 1) Operate each valve from the MCR 1) MCR display and local, visual remotely-operated valve can be and local control panel (if design observation indicate each valve operated remotely. has local valve control). fully opens and fully closes.

02. Verify each PSS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of air. 1) Isolate and vent air to the valve. fails to its safe position.

NuScale US460 SDAA 14.2-108 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-46: Test # 46 Process Sampling System (Continued)

03. Verify each PSS air-operated Place each valve in its non-safe 1) MCR display and local, visual valve fails to its safe position on position. observation indicate each valve loss of electrical power to its 1) Isolate electrical power to each fails to its safe position.

solenoid. air-operated valve.

04. Verify each PSS return pump to Align the PSS and CVCS to allow for 1) Local display and local, visual CVCS can be started and stopped return pump operation. observation indicate each return locally. 1) Start and stop each return pump pump starts and stops. Audible locally. and visible water hammer are not observed when each return pump starts.

05. Verify each PSS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each PSS on an MCS or PCS display, or is display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

46.02.XX System Level Tests 46.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify sampling capability of the 1) The NPM is in hot functional testing 1) The PSS analysis panel primary sampling points. with the RCS at normal operating instruments provide indication of pressure and the maximum the water analysis.

operating temperature achievable 2) The primary sampling ion by heating the RCS with the MHS. chromatography unit monitors for The RCS supply and discharge the programmed ion.

flow is in service. 3) An RCS injection flow grab sample Align the CVCS and PSS to is successfully obtained.

provide continuous sampling flow 4) An RCS discharge flow grab to the PSS analysis panel. sample is successfully obtained.

2) The RCS discharge line is in 5) A CVCS demineralizer discharge service. Align the RCS and PSS to flow grab sample is successfully provide sampling flow to the obtained.

primary sampling ion chromatography units.

3) Open the PSS grab sample panel manual valve to obtain an RCS injection flow pressurized grab sample.

4) Open the PSS grab sample panel manual valve to obtain an RCS discharge flow pressurized grab sample.

5) Open the PSS grab sample panel manual valve to obtain a CVCS demineralizer discharge flow pressurized grab sample.

NuScale US460 SDAA 14.2-109 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-46: Test # 46 Process Sampling System (Continued) 46.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify sampling capability of the The NPM is in hot functional testing 1) The PSS containment gas sample containment sampling points. with the RCS at normal operating panel instruments provide pressure and the maximum operating indication of the gas analysis.

temperature achievable by heating the RCS with the MHS.

The CES is in service.

1) Align the CES and PSS to provide continuous sampling flow to the PSS containment gas sample panel.

46.02.03.

Test Objective Test Method Acceptance Criteria

1. Verify sampling capability of the 1) The NPM is in hot functional testing 1) The PSS secondary sampling secondary sampling points. with the RCS at normal operating system FW and MS sample panel pressure and the maximum instruments provide indication of operating temperature achievable the water and steam analysis.

by heating the RCS with the MHS. 2) The FW and MS 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 3) The FW and MS ion to the PSS secondary sampling chromatography analysis panel system FW and MS sample panel. monitors the programmed ion.

2) Open the manual FW and MS ion 4) The FW and MS ion chromatography analysis panel chromatography analysis panel valve to obtain a FW to SG sample. monitors the programmed ion.

3) Open the manual FW and MS ion 5) The FW and MS ion chromatography analysis panel chromatography analysis panel valve to obtain an SG #1 steam monitors the programmed ion.

sample.

4) Open the manual FW and MS ion chromatography analysis panel valve to obtain an SG #2 steam sample.

5) Open the manual FW and MS ion chromatography analysis panel valve to obtain a condensate pump discharge sample.

NuScale US460 SDAA 14.2-110 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-47: Test # 47 High Voltage AC Electrical Distribution System Preoperational test is required to be performed once.

The EHVS is described in Section 8.3.1, 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 offsite nonsafety-related Component level tests transmission system by providing electrical power during normal operation and configuration management of utility.

47.00.XX Prerequisites

01. Verify an instrument calibration is performed on all EHVS instruments that provide information signals to the PCS for the bus and main power transformer under test.

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

03. Verify all protective devices associated with the EHVS bus and main power transformer under test are tested before that bus is energized, and approved test records indicate each protective device is calibrated within its required test interval.

47.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each EHVS breaker can be 1) Operate each breaker from the 1) MCR display and local, visual operated locally. local control panel while the observation indicate each breaker breaker is in the test position. opens and closes.

02. Verify each EHVS breaker can be 1) Operate each breaker from the 1) MCR display and local, visual operated remotely. MCR while the breaker is in the test observation indicate each breaker position. opens and closes.

03. Verify each EHVS breaker trips on 1) Simulate each fault condition for a 1) MCR display and local, visual its fault conditions. breaker when the breaker is in the observation indicate each breaker test position. opens on each fault condition.

04. Verify each EHVS bus can be 1) Energize each EHVS bus from its 1) Bus voltage is within design limits.

powered by offsite power via its main power transformer.

main power transformer.

(Test not required if an offsite power system is not provided.)

05. Verify each EHVS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each EHVS on an MCS or PCS display, or display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

47.02.XX System Level Tests None NuScale US460 SDAA 14.2-111 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-48: Test # 48 Medium Voltage AC Electrical Distribution System Preoperational test is required to be performed once. The testing of each EMVS bus that provides power to 00 loads (common system loads) is performed with the EMVS loads of the first NPM in power operation.

The EMVS is described in Section 8.3.1 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The EMVS supports the low nonsafety-related component-level tests voltage AC electrical distribution system by providing electrical power.

2. The EMVS supports the CHWS by nonsafety-related component-level tests providing electrical power to loads.

3. The EMVS supports the SCWS by nonsafety-related component-level tests providing electrical power to loads.

48.00.XX Prerequisites

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

02. Verify all protective devices associated with the EMVS bus and modular unit auxiliary transformer under test are tested before that bus is energized. Approved test records indicate each protective device is calibrated within its required test interval.

48.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each EMVS breaker can be 1) Operate each breaker from the 1) MCR display and local, visual operated locally. local control panel while the observation indicate each breaker breaker is in the test position. opens and closes.

02. Verify each EMVS breaker can be 1) Operate each breaker from the 1) MCR display and local, visual operated remotely. MCR while the breaker is in the test observation indicate each breaker position. opens and closes.

03. Verify each EMVS breaker trips on 1) Simulate each fault condition for a 1) MCR display and local, visual its fault conditions. breaker when the breaker is in the observation indicate each breaker test position. opens on each fault condition.

04. 1) 1) Bus voltage is within design limits.

a. Verify each EMVS bus can be a. Energize each EMVS bus from its powered via its modular unit modular unit auxiliary transformer.

auxiliary transformer. b. Energize each EMVS bus from an

b. Verify each EMVS bus can be adjacent EMVS bus.

powered via an adjacent bus.

(Test not required if an offsite power system is not provided.)

05. Verify each EMVS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each EMVS on an MCS or PCS display, or display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

06. Verify the automatic transfer of 1) Simulate all conditions that require 1) MCR display and local, visual each EMVS bus to each adjacent an automatic bus transfer to an observation indicate the required EMVS bus. adjacent bus. tie breaker from the adjacent bus Repeat for each adjacent EMVS bus. closes.

This test may be performed with the EMVS bus energized or deenergized.

48.02.XX System Level Tests None NuScale US460 SDAA 14.2-112 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-49: Test # 49 Low Voltage AC Electrical Distribution System Preoperational test is required to be performed in support of the testing of each NPM. The testing of each ELVS bus that provides power to common system loads is performed with the first NPM tested.

The ELVS is described in Section 8.3.1, 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.

49.00.XX Prerequisites

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

02. Verify all protective devices associated with the ELVS bus and station service transformer under test are tested before that bus is energized.

49.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each ELVS breaker can be 1) Operate each breaker from the 1) MCR display and local, visual operated locally. local control panel while the observation indicate each breaker breaker is in the test position. opens and closes.

02. Verify each ELVS breaker can be 1) Operate each breaker from the 1) MCR display and local, visual operated remotely. MCR while the breaker is in the test observation indicate each breaker position. opens and closes.

03. Verify each ELVS breaker trips on 1) Simulate each fault condition for a 1) MCR display and local, visual its fault conditions. breaker when the breaker is in the observation indicate each breaker test position. opens on each fault condition.

04. Verify each ELVS bus can be 1) Energize each ELVS bus from its 1) Bus voltage is within design limits.

powered by offsite power via its station service transformer.

station service transformer.

(Test not required if an offsite power system is not provided.)

05. Verify automatic bus transfer of Perform the following test for each of 1) The associated ELVS bus tie each ELVS bus. the ELVS buses. breaker closes.

1) Open the ELVS supply breaker to a given ELVS bus.

06. Verify each ELVS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each ELVS on an MCS or PCS display, or display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

49.02.XX System Level Tests None NuScale US460 SDAA 14.2-113 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-50: Test # 50 Augmented DC Power System Component level tests are required to be performed for each NPM, and once for the augmented DC power system (EDAS) common channels.

System Level Test 50.02.01 and Test 50.02.02 are required to be performed once. System Level Test 50.02.01 and Test 50.02.02 may be performed concurrently.

System Level Test 50.02.03 is required to be performed once for each NPM.

The EDAS is described in Sections 8.1.2, 8.1.3 and 8.3.2, and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The EDAS supports the following nonsafety-related All functions are verified by systems by providing DC electrical component-level tests.

power. System level tests provide additional

MPS verification as follows:

NMS 50.02.01

RMS 50.02.02

plant lighting system (PLS) 50.02.03

PPS

SDIS EDAS system functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

2. EDAS supports the MPS by nonsafety-related 59.01.07 providing EDAS module-specific operating parameter information signals.

3. EDAS supports the PPS by nonsafety-related 59.01.03 providing EDAS common operating parameter information signals.

50.00.XX Prerequisites

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

02. Verify a valve-regulated lead-acid battery acceptance tests is performed on all EDAS batteries to confirm battery capacity in accordance with Institue of Electrical and Electronics Engineers Standard 1188 Sections 6 and 7.

03. Verify battery charger performance testing is completed by the manufacturer or a site acceptance test is completed in accordance with manufacturer instructions.

50.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each EDAS bus can be 1) Configure the EDAS battery and 1) EDAS bus voltage is within design powered by its associated battery. battery charger(s) associated with limits.

an EDAS bus such that the battery is the only source of power to the bus.

Repeat the test for the remaining EDAS channels.

02. Verify each EDAS bus can be 1) Configure the EDAS battery and 1) EDAS bus voltage is within design powered by its associated battery battery charger(s) associated with limits.

charger(s). an EDAS bus such that a battery (Test may be performed as part of charger is the only source of power SAT.) to the bus.

2) Repeat the test if the bus has a standby battery charger.

Repeat the test for the remaining EDAS channels.

NuScale US460 SDAA 14.2-114 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-50: Test # 50 Augmented DC Power System (Continued)

03. Verify each EDAS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each EDAS on an MCS or PCS display, or display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

50.02.XX System Level Tests 50.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the EDAS common buses 1) With both EDAS common buses 1) The MCR lighting designed to be provide independent power to the energized and providing power to powered by the EDAS Division I MCR emergency lighting. MCR emergency lighting, common bus is de-energized, and (RG 1.41 Independence Test) de-energize the EDAS Division I the MCR emergency lighting common bus. designed to be powered by the

2) With both EDAS common buses EDAS Division II common bus is energized and providing power to energized.

MCR emergency lighting, 2) The MCR emergency lighting de-energize the EDAS Division II designed to be powered by the common bus. EDAS Division II common bus is de-energized, and the MCR emergency lighting designed to be powered by the EDAS Division I common bus is energized.

50.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify the EDAS common buses 1) With EDAS Division I and Division 1) Power is available in the MCR for provide independent power to all II common buses energized verify SDIS displays.

SDIS MCR displays. power is available in the MCR for 2)

(RG 1.41 Independence Test) all SDIS displays. a. Power is not available in the MCR

2) De-energize the EDAS Division I for SDIS Division I displays.

common bus. b. Power is available in the MCR for

3) Re-energize the EDAS Division I SDIS Division II displays.

common bus and de-energize the 3)

EDAS Division II common bus. a. Power is not available in the MCR for SDIS Division II displays.

b. Power is available in the MCR for SDIS Division I displays.

NuScale US460 SDAA 14.2-115 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-50: Test # 50 Augmented DC Power System (Continued) 50.02.03.

Test Objective Test Method Acceptance Criteria

1. Verify EDAS module-specific 1) With all EDAS module-specific 1) Power is available to the Division I channels provide independent and channels de-energized for the NPM ECCS trip valve solenoids.

redundant power to the ECCS trip under test, energize EDAS 2) valve solenoids and PAM Type B module-specific channel A. a. Power is available to the Division I and C variables. 2) With all EDAS module-specific ECCS trip valve solenoids.

(RG 1.41 Independence Test) channels de-energized for the NPM b. All PAM Type B and C variables under test, energize EDAS shown on Figure 7.1-2 are module-specific channel C. displayed on an SDIS display for

3) With all EDAS module-specific the NPM under test.

channels de-energized for the NPM 3) under test, energize EDAS a. Power is available to the Division II module-specific channel B. ECCS trip valve solenoids.

4) With all EDAS module-specific b. All PAM Type B and C variables channels de-energized for the NPM shown on Figure 7.1-2 are under test, energize EDAS Module displayed on an SDIS display for

-specific channel D. the NPM under test.

4) Power is available to the Division II ECCS trip valve solenoids.

NuScale US460 SDAA 14.2-116 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-51: Test # 51 Normal DC Power System Component-level tests 51.01.01 through 51.01.18 are required to be performed for the first NPM.

Component-level test 51.01.19 is required to be performed once per normal DC power system (EDNS) subsystem.

Component-level battery, battery charger, and inverter tests may be completed as part of SAT.

EDNS is described in Section Section 8.1.3 and Section 8.3.2 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The EDNS supports the following nonsafety-related Functions verified by prerequisite and systems by providing DC electrical component level tests.

power.

EHVS

EMVS

ELVS

TGS

2. The EDNS supports the following nonsafety-related Functions verified by prerequisite and systems by providing AC electrical component-level tests.

power.

communication system (COMS)

FDS

FPS

RMS

meteorological and environmental monitoring system

MCS

PCS

plant-wide video monitoring system

RBVS

seismic monitoring system (SMS)

TBVS 51.00.XX Prerequisites

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

02. Verify a valve-regulated lead-acid battery acceptance tests is performed on all EDNS batteries to confirm battery capacity in accordance with Institue of Electrical and Electronics Engineers Standard 1188 Sections 6 and 7.

03. Verify battery charger performance testing is completed by the manufacturer or a site acceptance test is completed in accordance with manufacturer instructions.

04. Verify inverter performance testing is completed by the manufacturer or a site acceptance test is completed in accordance with manufacturer instructions.

51.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each EDNS RXB 1) Configure the battery and battery 1) EDNS DC bus voltage is within subsystem DC bus can be charger associated with one of the design limits.

powered by its associated battery. EDNS RXB subsystems such that the battery is the only source of power to its associated DC bus.

Repeat the test for the other EDNS RXB subsystem.

NuScale US460 SDAA 14.2-117 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-51: Test # 51 Normal DC Power System (Continued)

02. Verify each EDNS RXB 1) Configure the battery and battery 1) EDNS DC bus voltage is within subsystem DC bus can be charger associated with one of the design limits.

powered by its associated battery EDNS RXB subsystems such that charger. the battery charger is the only source of power to its associated DC bus.

Repeat the test for the other EDNS RXB subsystem.

03. Verify each EDNS RXB 1) Energize the AC bus of one of the 1) EDNS AC bus voltage is within subsystem AC bus can be EDNS RXB subsystems from the design limits.

powered by its associated inverter source of its associated inverter. inverter.

Repeat the test for the other EDNS RXB subsystem.

04. Verify each EDNS RXB 1) Energize the AC bus of one of the 1) EDNS AC bus voltage is within subsystem AC bus can be EDNS RXB subsystems from the design limits.

powered by its associated voltage voltage regulating transformer regulating transformer. source of its associated inverter.

Repeat the test for the other EDNS RXB subsystem.

05. Verify the EDNS CRB subsystem 1) Configure the battery and battery 1) EDNS DC bus voltage is within DC bus can be powered by its charger associated with the EDNS design limits.

associated battery. CRB subsystem such that the battery is the only source of power to its associated DC bus.

06. Verify the EDNS CRB subsystem 1) Configure the battery and battery 1) EDNS DC bus voltage is within DC bus can be powered by its charger associated with the EDNS design limits.

associated battery charger. CRB subsystem such that the battery charger is the only source of power to its associated DC bus.

07. Verify each EDNS CRB 1) Energize the EDNS CRB 1) EDNS AC bus voltage is within subsystem AC bus can be subsystem AC bus from the design limits.

powered by its associated inverter source of its associated inverter. inverter.

08. Verify each EDNS CRB 1) Energize the EDNS CRB 1) EDNS AC bus voltage is within subsystem AC bus can be subsystem AC bus from the design limits.

powered by its associated voltage voltage regulating transformer regulating transformer. source of its associated inverter.

09. Verify the EDNS RWB subsystem 1) Configure the battery and battery 1) EDNS DC bus voltage is within DC bus can be powered by its charger associated with the EDNS design limits.

associated battery. RWB subsystem such that the battery is the only source of power to its associated DC bus.

10. Verify the EDNS RWB subsystem 1) Configure the battery and battery 1) EDNS DC bus voltage is within DC bus can be powered by its charger associated with the EDNS design limits.

associated battery charger. RWB subsystem such that the battery charger is the only source of power to its associated DC bus.

11. Verify each EDNS RWB 1) Energize the EDNS RWB 1) EDNS AC bus voltage is within subsystem AC bus can be subsystem AC bus from the design limits.

powered by its associated inverter source of its associated inverter. inverter.

12. Verify each EDNS RWB 1) Energize the EDNS RWB 1) EDNS AC bus voltage is within subsystem AC bus can be subsystem AC bus from the design limits.

powered by its associated voltage voltage regulating transformer regulating transformer. source of its associated inverter.

NuScale US460 SDAA 14.2-118 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-51: Test # 51 Normal DC Power System (Continued)

13. Verify each EDNS PDC 1) Configure the battery and battery 1) EDNS DC bus voltage is within subsystem DC bus can be charger associated with the EDNS design limits.

powered by its associated battery. PDC subsystems such that the battery is the only source of power to its associated DC bus.

Repeat the test for all buses of EHV PDC, EMV PDC, and ELV PDC.

14. Verify each EDNS PDC 1) Configure the battery and battery 1) EDNS DC bus voltage is within subsystem DC bus can be Charger associated with the EDNS design limits.

powered by its associated battery PDC subsystem such that the charger. battery charger is the only source of power to its associated DC bus.

2) Repeat the test for the swing battery charger if the bus has a swing battery charger.

Repeat the test for all buses of EHV PDC, EMV PDC, and ELV PDC.

15. Verify each EDNS PDC 1) Energize the EDNS PDC 1) EDNS AC bus voltage is within subsystem AC bus can be subsystem AC bus from the design limits.

powered by its associated inverter source of its associated inverter. inverter.

Repeat the test for all buses of EHV PDC, EMV PDC, and ELV PDC.

16. Verify each EDNS PDC 1) Energize the EDNS PDC 1) EDNS AC bus voltage is within subsystem AC bus can be subsystem AC bus from the design limits.

powered by its associated voltage voltage regulating transformer regulating transformer. source of its associated inverter.

Repeat the test for all buses of EHV PDC, EMV PDC, and ELV PDC.

17. Verify each EDNS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each EDNS on an MCS or PCS display, or display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display) 51.02.XX System Level Tests None NuScale US460 SDAA 14.2-119 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-52: Test # 52 Backup Power Supply System Preoperational test is required to be performed once.

The BPSS is described in Section 8.3.1, and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The BPSS supports EMVS by nonsafety-related 52.02.01 providing diesel generator backup electrical power.

52.00.XX Prerequisites

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

02. Verify all protective devices associated with the BPSS diesel generators have been tested before performing this test.

52.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each BPSS breaker can be 1) Operate each breaker from the 1) MCR display and local, visual operated locally. local control panel while the observation indicate each breaker breaker is in the test position. opens and closes.

02. Verify each BPSS breaker can be 1) Operate each breaker from the 1) MCR display and local, visual operated remotely. MCR while the breaker is in the test observation indicate each breaker position. opens and closes.

03. Verify the BPSS diesel generators Align the BPSS to allow for diesel 1) MCR display and local, visual can be started and stopped locally generator operation. observation indicate the diesel and remotely. 1) Start and stop the diesel generator generator started and stopped.

from the MCR. 2) MCR display and local, visual

2) Start and stop the diesel generator observation indicate the diesel locally. generator started and stopped.

Repeat the test for the other diesel generator.

04. Verify the BPSS diesel generator Align a fuel oil transfer pump to provide 1) MCR display and local, visual day tank fuel oil transfer pumps oil to its associated day tank. observation indicate the transfer automatically maintain day tank 1) Simulate a low level in the day pump starts and then stops when levels. tank. day tank level reaches the high Repeat the test for each day tank fuel level setpoint.

oil transfer pump.

05. Verify each BPSS instrument is 1) Initiate a single real or simulated 1) The instrument signal is displayed available on an MCS or PCS instrument signal from each BPSS on an MCS or PCS display, or display. transmitter. recorded by the applicable control (Test not required if the instrument system historian.

calibration verified the MCS or PCS display.)

52.02.XX System Level Tests 52.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify BPSS diesel generator Align the BPSS to allow for diesel 1) MCR display and local, visual automatically starts and achieves generator operation. observation indicate the diesel rated voltage and frequency. 1) Initiate a real or simulated loss of generator started and achieved power signal. rated voltage and frequency.

NuScale US460 SDAA 14.2-120 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-53: Test # 53 Plant Lighting System Preoperational test is required to be performed once.

The 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 nonsafety-related 53.01.01 providing normal lighting.

2. The PLS supports the CRB by nonsafety-related 53.01.02 providing emergency lighting in the MCR.

3. The PLS supports the RXB by nonsafety-related 53.01.01 providing normal lighting.

4. The PLS supports the RXB by nonsafety-related 53.01.03 providing emergency lighting for post-fire safe-shutdown activities outside of the MCR and RSS.

53.00.XX Prerequisites N/A (Note: Component level test 53.01.03. supports ITAAC and the requirements of NFPA 804.)

53.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify the PLS provides normal 1) With normal MCR lighting in 1) illumination of the MCR operator service, measure the light at each a. The PLS provides lighting levels workstations, and the MCR safety MCR workstation. for a computer-based control room display information panel. specified in NUREG-0700, Revision 3.

02. The PLS provides emergency 1) With MCR emergency illumination 1) illumination of the MCR operator in service, measure the light at a. The PLS provides at least 10 workstations and the MCR safety each MCR workstation and MCR foot-candles of illumination at the display information panel. safety display information panel. MCR operator workstations and the MCR safety display information panel.

[ITAAC 03.08.02]

03. Verify the eight-hour battery pack 1) With no AC power available, 1) The required target areas are emergency lighting fixtures measure the light at each illuminated to provide at least one provide illumination for post-fire eight-hour battery pack emergency foot-candle illumination in the safe-shutdown activities lighting fixture target area. areas outside the MCR where performed by operators outside post-fire safe-shutdown activities the MCR. are performed.

[ITAAC 03.08.01]

53.02.XX System Level Tests None NuScale US460 SDAA 14.2-121 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-54: Test # 54 Module Control System 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.

On-site testing of the system is performed by module control system SAT.

The MCS is a distributed control system that 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 for nonsafety displays in the MCR and other locations where MCS human-system interfaces 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 is conducted. This FAT is conducted in accordance with a written test procedure for testing the software and hardware of the MCS before installation in the plant. Following installation, SAT must 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 MCR. 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 available on an MCS or PCS display. These component-level tests are described in the test abstracts of the systems that contain the instrument.

54.00.XX Prerequisites

01. Prerequisites associated with MCS testing are identified in the test abstracts that contain module-specific components that ensure communication with the MCS.

54.01.XX Component Level Tests None 54.02.XX System Level Tests None NuScale US460 SDAA 14.2-122 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-55: Test # 55 Plant Control System 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.

On-site testing of the system is performed by PCS SAT.

The PCS is a distributed control system that allows monitoring and control of virtually all module-specific plant components. The PCS includes all manual controls and video display units (VDUs) 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 VDUs for nonsafety displays in the MCR, 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, bi-directional interface with the MCS. The network interface devices for the PCS domain controller and 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 is conducted. This FAT is conducted in accordance with a written test procedure for testing the software and hardware of the PCS before installation in the plant. Following installation, SAT must 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 MCR. 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 available on an MCS or PCS display. These component-level tests are described in the test abstracts of the systems that contain the instrument.

55.00.XX Prerequisites

01. 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.

55.01.XX Component Level Tests None 55.02.XX System Level Tests None NuScale US460 SDAA 14.2-123 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-56: Test # 56 Module Protection System 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 and power ascension testing are:

System Function System Function Categorization Function Verified by Test #

1. The MPS supports the CNTS by safety-related 56.02.04 removing electrical power to the trip solenoids of the following CIVs on a CNTS isolation actuation signal:

RCS injection CIVs

RCS discharge CIVs

PZR spray CIVs

RPV high point degasification CIVs

Feedwater isolation valves (FWIVs)

Main steam isolation valves (MSIVs)

Main steam isolation bypass valves (MSIBV)

Containment evacuation system CIVs

Reactor component cooling water system CIVs

CFDS CIVs

2. The MPS supports the CNTS by safety-related 56.02.04 removing electrical power to the trip solenoids of the following valves on a DHRS actuation signal.

MSIVs

MSIBV

FWIVs

3. The MPS supports the ECCS by safety-related 56.02.04 removing electrical power to the trip solenoids of the following valves on an ECCS actuation signal.

RVVs

RRVs

4. The MPS supports the CNTS by safety-related 56.02.04 removing electrical power to the trip solenoids of the following CIVs on a CVCS isolation actuation signal:

RCS injection CIVs

RCS discharge CIVs

PZR spray CIVs

RPV high point degasification CIVs

5. The MPS supports the CVCS by safety-related 56.02.04 removing electrical power to the trip solenoids of the DWS supply isolation valves on a DWS isolation actuation signal.

NuScale US460 SDAA 14.2-124 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-56: Test # 56 Module Protection System (Continued)

6. The MPS supports the ECCS by safety-related 56.02.04 removing electrical power to the trip solenoids of the reactor vent valves on an LTOP actuation signal.

7. The MPS supports the ELVS by safety-related 56.02.06 removing electrical power to the PZR heaters on a PZR heater trip actuation signal.

8. The MPS supports the ELVS by safety-related 56.02.06 removing electrical power to the CRDS for a reactor trip.

9. The DHRS supports the RCS by safety-related 56.02.04 opening the DHRS actuation valves on a DHRS actuation signal for DHRS operation.

10.The CNTS supports the DHRS by safety-related 56.02.04 closing CIVs for the MS and FW systems when actuated by the MPS.

11. The CNTS supports the RCS by safety-related 56.02.04 closing the CIVs for PZR spray, RCS injection, RCS letdown, and RPV high point degasification when actuated by the MPS.

12. The CNTS supports the RXB by safety-related 56.02.04 providing a barrier to contain mass, energy, and fission product release by closure of the CIVs upon a containment isolation signal.

13. The ECCS supports the RCS by safety-related 56.02.02 opening the ECCS RVVs and RRVs when their respective trip valve is actuated by the MPS.

14. The ECCS supports the RCS by safety-related 56.02.02 providing recirculated coolant from the containment to the RPV for the removal of core heat.

15. The ECCS supports the RCS by safety-related 56.02.02 providing LTOP for maintaining the reactor coolant pressure boundary.

16. The CVCS supports the RCS by safety-related 56.02.04 isolating dilution sources.

17. The FWS supports the CNTS by nonsafety-related 56.02.04 providing secondary isolation of the FW lines.

18. The MSS supports the CNTS by nonsafety-related 56.02.04 providing secondary isolation of the MS lines.

19. The FWS supports the DHRS by nonsafety-related 56.02.04 providing secondary isolation of the FW lines, ensuring required boundary conditions for DHRS operation.

NuScale US460 SDAA 14.2-125 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-56: Test # 56 Module Protection System (Continued)

20. The NMS supports the MPS by safety-related 56.02.01 providing neutron flux data for 86.03.01 various reactor trips.

21. ECCS supports MPS by providing nonsafety-related 56.02.01 instrumentation information signals.

22. The DHRS supports the MPS by safety-related 56.02.01 providing MPS actuation instrument information signals.

23. The RCS supports the MPS by nonsafety-related 56.02.01 providing instrument information signals.

24. The RCS supports the MPS by safety-related 56.02.01 providing instrument information signals for LTOP actuation.

25. The MPS supports the DHRS by safety-related 56.02.04 removing electrical power to the trip solenoids of the DHRS actuation valves on a DHRS actuation signal.

26. The MPS supports the CNTS by safety-related 56.02.01 providing power to sensors.

27. The MPS supports the DHRS by safety-related 56.02.01 providing power to sensors.

28. The MPS supports the RCS by safety-related 56.02.01 providing power to sensors.

56.00.XX Prerequisite

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

56.01.XX Component Level Tests None 56.02.XX System Level Tests 56.02.01.

Test Objective Test Method Acceptance Criteria

1. Verify the instrument signals of Table 7.1-2 lists all of sensors that 1) Each MPS monitored signal is MPS monitored variables are input to MPS. displayed on an MCR workstation displayed in the MCR. This test may be performed and the module-specific safety concurrently with SDIS test 59.02.02 display instrument panel (if for PAM Type B and Type C testing designed for safety display described in Section 14.2.12 instrument display).

1) Inject a single signal as close as practicable 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 is necessary to test the signal from the sensor to the disconnect and then from the disconnect to the MCR display.

NuScale US460 SDAA 14.2-126 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-56: Test # 56 Module Protection System (Continued) 56.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify each ECCS reactor vent This test verifes the stroke time of 1) Each ECCS reactor recirculation valve and reactor recirculation each RRV and RVV and verifes ECCS valve and reactor vent valve travels valve operates to satisfy its engineered safety feature actuation from fully closed to fully open in ESF-actuated design stroke time. capability from the MCR by actuating less than or equal to the time

2. Verify the MPS can manually the valves with RCS pressure below specified in TS.

actuate ESF equipment from the the IAB low RCS pressure threshold. 2) The MPS actuates the ESF MCR. 1) Verify all RVVs and RRVs are equipment to perform its

3. Verify deliberate operator action is closed. safety-related function as required to return the ESF 2) Initiate a manual ECCS described in Table 7.1.4.

actuated equipment to its engineered safety feature [ITAAC 02.01.14]

non-actuated position. actuation signal from the MCR. [ITAAC 02.01.19]

3) [ITAAC 02.05.06]

a. Attempt to operate ECCS from the [ITAAC 02.05.07]

MCR. [ITAAC 02.05.08]

b. Remove the manual ESF actuation 3) signal and attempt to operate a. ECCS cannot be operated from the ECCS from the MCR. MCR.

c. Use the MCR enable nonsafety b. ECCS cannot be operated from the control switch to allow operation of MCR.

ECCS from the MCR. c. ECCS can be operated from the Repeat for LTOP engineered safety MCR.

feature actuation [ITAAC 02.01.14]

[ITAAC 02.05.08]

56.02.03.

Test 56.02.03 is performed concurrently with Test 56.02.04, which operates all of the ESF actuation valves during hot functional testing.

Test 56.02.03. records the stroke times of DHRS actuation valves as they travel to their ESF-actuated position with the RCS pressure at normal operating pressure.

Test Objective Test Method Acceptance Criteria

1. Verify each DHRS actuation valve 1) Time the operation of all DHRS 1) Each DHRS actuation valve travels operates to satisfy its actuation valves as they actuate to from fully closed to fully open in ESF-actuated design stroke time. their ESF position during the less than or equal to the time manual ESF actuation testing in specified in TS.

Test 56.02.04.

NuScale US460 SDAA 14.2-127 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-56: Test # 56 Module Protection System (Continued) 56.02.04.

Test 56.02.04 is performed at hot functional testing concurrently with Turbine Generator System Test 29.02.01 to allow testing of ESF actuations at normal operating pressure and elevated temperatures. Test 29.02.01 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 before fuel load.

Test Objective Test Method Acceptance Criteria

1. Verify the MPS can manually Table 7.1-4 lists all of the ESF 1) The MPS actuates the ESF actuate ESF equipment from the functions. equipment to perform its MCR. The RCS is at normal operating safety-related function as

2. Verify deliberate operator action pressure supplying bypass steam to described in Table 7.1-4.

is required to return the ESF the condenser. [ITAAC 02.01.13]

actuated equipment to its 1) Initiate a manual ESF actuation [ITAAC 02.01.15]

non-actuated position. signal from the MCR. [ITAAC 02.01.18]

3. Verify no dynamic effects caused 2) [ITAAC 02.01.20]

by changes in fluid flow. a. Attempt to operate the actuated [ITAAC 02.05.06]

ESF equipment from the MCR. [ITAAC 02.05.07]

b. Remove the manual ESF actuation [ITAAC 02.05.08]

signal and attempt to operate the 2) actuated ESF equipment from the a. The actuated equipment cannot be MCR. operated from the MCR.

c. Use the MCR enable nonsafety b. The actuated equipment cannot be control switch to allow operation of operated from the MCR.

the ESF actuated equipment from c. The ESF equipment can be the MCR. operated from the MCR.

Repeat as necessary to ensure the [ITAAC 02.01.13]

following ESF functions are tested: [ITAAC 02.01.15]

DHRS [ITAAC 02.05.08]

secondary system isolation 3)

CNTS isolation a. Audible indications of water

DWS isolation hammer are not observed.

CVCS isolation b. No damage to pipe supports or

PZR heater trip restraints.

c. No damage to equipment.

d. No equipment leakage.

56.02.05.

Test 56.02.05 is performed concurrently with Test 56.02.04, which operates all of the ESF actuation valves during hot functional testing.

Test 56.02.05 records the stroke times of CIVs as they travel to their ESF-actuated position with the RCS pressure at normal operating pressure.

Test Objective Test Method Acceptance Criteria

1. Verify the CIVs operate to satisfy Table 6.2-4 contains the design 1) Each CIV travels from fully open to their ESF-actuated design stroke closure time for containment isolation fully closed in less than or equal to time. valves. the time listed in Table 6.2-4 after

1) Time the operation of all CIVs as receipt of a containment isolation they actuate to their ESF position signal.

during the manual ESF actuation [ITAAC 02.01.08]

testing in 56.02.04.

NuScale US460 SDAA 14.2-128 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-56: Test # 56 Module Protection System (Continued) 56.02.06.

This test verifes 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 PZR heater supply breaker.

Test Objective Test Method Acceptance Criteria Verify the MPS response times from Section 7.1.4 contains a description of 1) The MPS reactor trip functions sensor output through: design-basis event actuation delays listed in Table 7.1-3 and ESF

1. Reactor trip breaker actuation for assumed in the plant safety analysis functions listed in Table 7.1-4 have the reactor trip function. and listed in Table 7.1-6. The response times that are less than

2. De-energization of the associated actuation delays do not include ESF or equal to the design-basis safety solenoid valve for ESF-actuated actuated component delays for analysis response time valves. actuated valves. assumptions in Table 7.1-6.

3. Opening of the PZR heater supply 1) Perform a time response test for [ITAAC 02.05.09]

breaker for the PZR heater trip. the actuation signals listed in Table 7.1-6.

56.02.07.

Test Objective Test Method Acceptance Criteria

1. Verify MCR alarms when The purpose of this test is to verify 1) Each automatic operating bypass automatic operating bypasses are MCR alarms, not to verify the logic of is alarmed in the MCR.

established. the operating bypasses. Any signal [ITAAC 02.05.10]

2. Verify MCR alarms when manual that establishes the bypass can be 2) Each manual operating bypass is operating bypasses are used. alarmed in the MCR.

established. Table 7.1-5 contains a list of operating [ITAAC 02.05.10]

3. Verify MPS maintenance bypasses. 3) The inoperable status of the SFM bypasses are indicated in the 1) For automatically established is provided in the MCR.

MCR. operating bypasses perform the [ITAAC 02.05.11]

following:

a. Simulate the logic required to establish the operating bypass.

b. Remove the logic.

c. Repeat for each automatically established operating bypass.

2) For manually established operating bypasses perform the following:

a. Simulate the logic required to allow the operating bypass to be established.

b. Manually establish the operating bypass.

c. Repeat the logic.

d. Repeat for each manually established operating bypass.

3)

a. Place a safety function module (SFM) in maintenance bypass by using the out of service and trip/bypass switches associated with the SFM.

b. Repeat tests for all SFMs.

NuScale US460 SDAA 14.2-129 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-57: Test # 57 Plant Protection System Preoperational test is required to be performed once.

The PPS is described in Section 7.0.4. The 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 15.02.01 providing actuation and control signals to the CRE isolation dampers.

2. The PPS supports the CRHS by nonsafety-related 15.02.01 providing actuation and control signals.

3. The PPS supports the CRVS by nonsafety-related 16.02.03 providing actuation and control signals to the outside air isolation dampers.

57.00.XX Prerequisites

01. Verify an instrument calibration is completed, with approved records and calibration due dates, for instruments required to perform this test.

57.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each variable monitored by 1) Initiate a single real or simulated 1) Each PPS variable is displayed on PPS is available on an MCS or instrument signal from each an MCS or PCS display, or is PCS display. transmitter monitored by PPS. recorded by the applicable control system historian.

57.02.XX System Level Tests None NuScale US460 SDAA 14.2-130 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-58: Test # 58 Neutron Monitoring System Preoperational test is required to be performed for each NPM.

The neutron monitoring system (NMS) is described in Section 7.0.4. NMS functions are not verified by NMS tests. NMS functions verified by other tests are:

System Function System Function Categorization Function Verified by Test #

1. The NMS supports the MPS by safety-related 56.02.01 providing neutron flux data for 86.03.01 various reactor trips.

2. The NMS supports the MPS by nonsafety-related 59.02.02 providing information signals for PAM.

3. The NMS supports the MPS by nonsafety-related 59.02.02 providing information signals for PAM during containment vessel flooded conditions.

58.00.XX Prerequisites

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

58.01.XX Component Level Tests None 58.02.XX System Level Tests None NuScale US460 SDAA 14.2-131 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-59: Test # 59 Safety Display and Indication System Component-level testing for the module-specific SDIS is required to be performed for each NPM.

Component-level testing for the common SDIS is required to be performed once.

Test 59.02.01 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 59.02.02 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 and the functions verified by this test are:

System Function System Function Categorization Function Verified by Test #

1. The SDIS actively supports the nonsafety-related Module-specific SDIS CRB by providing the MCR component-level tests accident monitoring plant Common SDIS component-level tests conditions. 59.02.01 59.02.02

2. The SDIS actively supports the nonsafety-related Module-specific SDIS PCS by providing plant status and component-level tests indication data to the plant data Common SDIS component-level tests historian. 59.02.01 59.02.02

3. The ICIS supports the MPS by nonsafety-related 59.02.02 providing RXCS temperature information.

4. The ECCS supports MPS by nonsafety-related 59.02.02 providing PAM instrument information signals.

5. The RCS supports the MPS by nonsafety-related 59.02.02 providing PAM instrument information signals.

6. The CNTS supports the MPS by nonsafety-related 59.02.02 providing PAM information signals.

7. The RMS supports the RXB by nonsafety-related 59.02.02 monitoring radiation levels in the building in proximity of the bioshield.

8. The NMS supports the MPS by nonsafety-related 59.02.02 providing information signals for PAM.

9. The NMS supports the MPS by nonsafety-related 59.02.02 providing information signals for PAM during containment vessel flooded conditions.

10. The decay heat removal system nonsafety-related 59.02.02 supports the MPS by providing PAM instrument information signals.

11. The EDAS supports the PPS by nonsafety-related 59.01.03 providing common EDAS operating parameter information signals.

12. The EDAS supports the MPS by nonsafety-related 59.01.07 providing module-specific EDAS operating parameter information signals.

NuScale US460 SDAA 14.2-132 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-59: Test # 59 Safety Display and Indication System (Continued) 59.00.XX Prerequisites

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

Note:

Testing of PAM Type B and Type C displays and alarms is performed in 59.02.01.

Note:

Testing of NPM level, pressure, and temperature and flow instruments is performed in 59.02.02.

59.01.XX Component Level Tests: Common SDIS Test Test Objective Test Method Acceptance Criteria

01. Verify the proper valve position 1) Open and close each valve 1) The valve opens and closes as indication for each valve that monitored by PPS. indicated by a common SDIS provides input to the PPS. display and an MCR workstation display.

02. Verify radiation monitor indication 1) Provide a simulated signal for each 1) The radiation signal is displayed is obtained in the MCR for each radiation monitor monitored by by a common SDIS display and an radiation monitor that provides PPS. MCR workstation.

input to the PPS.

03. Verify EDAS and ELVS voltage 1) Provide a simulated signal for each 1) The voltage signal is displayed by indication is obtained in the MCR EDAS and ELVS voltmeter a common SDIS display and an for voltmeters that provide input to monitored by PPS. MCR workstation.

the PPS.

04. Verify instrument indication is 1) Provide a simulated signal for each 1) The instrument signal is displayed obtained in the MCR for instrument monitored by PPS. by a common SDIS display and an instruments that provide input to MCR workstation.

the PPS.

Component Level Tests: Module Specific SDI Test Test Objective Test Method Acceptance Criteria

05. Verify the proper valve position 1) With the NPM assembled, open 1) The valves open and close as indication for each ESF valves and close the valves listed in indicated by a module-specific that provide input to MPS. Table 7.1-2. SDIS display and an MCR

2) Provide a real or simulated signal workstation display.

for each reactor safety valve 2) The valve opens and closes as position (Table 7.1-2). indicated by a module-specific SDIS display and an MCR workstation display.

06. Verify radiation monitor indication 1) Provide a simulated signal for each 1) The radiation monitor signal is is obtained in the MCR for each radiation monitor monitored by displayed by a module-specific radiation monitor that provides MPS listed in Table 7.1-2. SDIS display and an MCR input to the MPS. workstation.

07. Verify EDAS and ELVS voltage 1) Provide a simulated signal for each 1) The voltage signal is displayed by indication is obtained in the MCR EDAS and ELVS voltmeter a module-specific SDIS display for each voltmeter that provide monitored by MPS (Table 7.1-2). and an MCR workstation.

input to the MPS.

08. Verify neutron flux indication is 1) Provide a simulated signal for each 1) The neutron flux signal is obtained in the MCR for each neutron flux instrument monitored displayed by a module-specific radiation monitor that provides by MPS (Table 7.1-2). SDIS display and an MCR input to the MPS. workstation display.

09. Verify a neutron flux instrument 1) Provide a simulated signal for each 1) The neutron flux instrument fault is fault indication is obtained in the neutron flux instrument fault displayed by a module-specific MCR for each signal that provides monitored by MPS (Table 7.1-2). SDIS display and an MCR input to the MPS. workstation display.

NuScale US460 SDAA 14.2-133 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-59: Test # 59 Safety Display and Indication System (Continued) 59.02.XX System Level Test 59.02.01.

Test 59.02.01 is conducted concurrently with Turbine Generator System Test 29.02.01, which warms the RCS from ambient conditions to the highest temperature achievable by MHS heating.

Test Objective Test Method Acceptance Criteria

1. Verify that the output signals from 1) Increase RCS temperature from 1) All instruments track within the NPM level, pressure, ambient to the highest temperature acceptable design limits.

temperature, and flow instruments achievable by MHS heating. (Use TS channel check limits, when listed in Table 7.1-2 properly trend 2) Using the MCS, historian records applicable).

while increasing RCS temperature the engineering values for the and pressure. output of the instruments described Note: This is not a verification of in the test objective. Record data at instrument calibrations. 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 MCR workstations.

59.02.02.

Test Objective Test Method Acceptance Criteria

1. Verify PAM Type B and C 1) Simulate an injection signal for the 1) The PAM Type B and C variables variables are displayed on the PAM Type B and C variables listed listed in Table 7.1-7 are retrieved module-specific SDIS displays in in Table 7.1-7. and displayed on the SDIS the MCR. 2) Increase or decrease a simulated displays in the MCR.

2. Verify alarms associated with injection signal for the PAM Type B [ITAAC 02.05.13]

PAM Type B and C variables are and C variables listed in 2) The alarms associated with the retrieved in the MCR. Table 7.1-7 to obtain its associated PAM Type B and C variables listed

3. Verify module-specific PAM Type alarm. in Table 7.1-7 are retrieved and D variables are displayed on the 3) Simulate an injection signal for the displayed on the SDIS displays in module-specific SDIS displays in PAM Type D variables listed in the MCR.

the MCR. Table 7.1-7. 3) The PAM Type D variables listed in Table 7.1-7 are retrieved and displayed on the SDIS displays in the MCR.

NuScale US460 SDAA 14.2-134 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-60: Test # 60 Fixed-Area Radiation Monitoring System Preoperational test is required to be performed once.

The fixed-area radiation monitoring system (RMS) is described in Section 12.3.4 and the function verified by this test is:

System Function System Function Categorization Function Verified by Test #

1. The RMS supports the following nonsafety-related Component-level test 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 #

2. The RMS supports the RXB by nonsafety-related 59.02.02 monitoring radiation levels in the building in proximity of the bioshield.

60.00.XX Prerequisites

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

60.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify each fixed airborne 1) Actuate the check source on a MCR display and local, visual radiation monitor's response to an fixed airborne radiation monitor observation indicate the following:

alarm condition. listed in Table 12.3-9. 1) The MCR audible and visual Repeat test for the remainder of fixed alarms are received.

airborne radiation monitors. 2) The local readout, audible alarm, and visual alarm are received.

02. Verify each fixed area radiation 1) Actuate the check source on a MCR display and local, visual monitor's response to an alarm fixed area radiation monitor listed observation indicate the following:

condition. in Table 12.3-10. Repeat test for 1) The MCR audible and visual the remainder of fixed area alarms are received.

radiation monitors. 2) The local readout, audible alarm, and visual alarm are received.

60.02.XX System Level Tests None NuScale US460 SDAA 14.2-135 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-61: Test # 61 Communication System Preoperational test is required to be performed after construction turnover of the 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 #

1. The COMS supports the following nonsafety-related 61.01.01 locations by providing voice and 61.01.02 data communications within the 61.01.03 building and surrounding areas. 61.01.04

RXB

TGB

RWB

CRB

Security Buildings

ANB

Diesel Generator Building

Administrative and Training Building

CUB

Warehouse Building

Fire Water Building

Switchyard

Site plant cooling structures

Site water intake/discharge structure

Site utility rack structure 61.00.XX Prerequisites

01. Required communication system SAT have been completed and approved.

61.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify the wide area mass Station test personnel in each required 1) The test announcement is heard at notification system can be heard test area of the plant to monitor the each test site.

throughout the plant site. wide area mass notification system. 2) The test emergency alarm is heard

1) Use the public address to provide a at each test site.

test announcement.

2) Use the general alarm system to provide a test alarm.

02. Verify plant radio communications 1) Station test personnel in each 1) The plant radio communication is can be heard throughout the plant required test area of the plant to obtained at each test site.

site. communicate using plant radios.

03. Verify wireless communication 1) Station test personnel in each 1) The voice and data throughout the plant site. required test area of the plant to communication is obtained at each communicate using voice and data test site.

communication.

04. Verify the central alarm station is 1) Test the conventional (landline) 1) The conventional service connects equipped with a conventional service from the central alarm with the MCR and the local law (landline) telephone service that station to the MCR and local law enforcement authorities.

can be used to communicate with enforcement authorities. [ITAAC 03.16.11]

the MCR and local law enforcement authorities.

NuScale US460 SDAA 14.2-136 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-61: Test # 61 Communication System (Continued)

05. Verify that plant radio 1) Test communications with the plant 1) The radios must provide communications maintains radio system in areas described in continuous communications in all continuous communications the physical protection program test areas.

among the central alarm station boundaries and areas described in [ITAAC 03.16.12]

and on-duty watchmen, armed the contingency response event security officers, armed areas.

responders, or other security personnel who have responsibilities within the physical protection program and during contingency response events.

06. Verify all nonportable 1) Remove normal power from the 1) The nonportable communication communication devices (including central alarm station nonportable devices establish connections with conventional telephone systems) communication devices. the normal power removed.

in the central alarm station remain [ITAAC 03.16.13]

operable during the loss of normal power.

61.02.XX System Level Tests None NuScale US460 SDAA 14.2-137 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-62: Test # 62 Seismic Monitoring System The SMS is described in Section 3.7.4.

COL Item 14.2-6: An applicant that references the NuScale Power Plant US460 standard design will provide a test abstract for the seismic monitoring system preoperational testing.

System Function System Function Categorization Function Verified by Test #

1. As described in Section 3.7.4. nonsafety-related Provided by applicant.

62.00.XX Prerequisites Provided by applicant.

62.01.XX Component Level Tests Provided by applicant.

62.02.XX System Level Tests Provided by applicant.

NuScale US460 SDAA 14.2-138 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Test # 63 Hot Functional Testing 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 Test Objective Verified by Test # Tested Function Tests Categorization

1. CES 1) Verifies the automatic 1) 36.02.01 nonsafety-related operation of the CES to 2) 36.02.02 establish and maintain 3) 36.02.03 design vacuum for the containment vessel.

2) Verify radiation isolation and flow diversion on high radiation level in the CES.

3) Verifies the CES supports RCS leakage detection.

2. CNTS 1) Verifies each CNTS 1) 38.02.02 safety-related safety-related check valves open and close under preoperational conditions.

3. CVCS 1) Verifies CVCS automatic 1) 33.02.01 nonsafety-related operation to maintain 2) 33.02.02 PZR level. 3) 33.02.03

2) Verifies automatic PZR pressure control.

3) Verifies CVCS automatic boration and dilution of the RCS.

4. ECCS 1) Each ECCS valve opens 1) 40.02.01 safety-related after receipt of an ESF signal and after RCS pressure is decreased to the threshold pressure for operation of the IAB

5. FW system 1) Verifies the FWS 1) 29.02.01 nonsafety-related automatically controls 2) 29.02.01 flow to the SGs to maintain SG inventory.

2) Verifies the FWS automatically cools the turbine generator bypass steam flow in the MS desuperheater.

6. ICIS 1) Verifies proper 1) 42.02.01 nonsafety-related temperature indication is obtained from the ICIS thermocouples.

7. LRWS 1) Verifies the LRWS 1) 30.02.01 nonsafety-related receives and processes a gaseous stream from the PZR.

NuScale US460 SDAA 14.2-139 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Test # 63 Hot Functional Testing (Continued)

8. MHS 1) Verifies the MHS is 1) 29.02.01 nonsafety-related capable of heating the 2) 29.02.01 RCS to a temperature 3) 29.02.01 sufficient to obtain criticality.

2) Verifies the MHS is capable of heating the RCS to establish natural circulation flow sufficient to obtain criticality.

3) Verifies a local grab sample can be obtained from an MHS grab sample device indicated on the MHS piping and instrumentation diagram.

9. MPS 1) Verifies design 1) 56.02.04 safety-related responses to manual 2) 56.02.05 ESF signals.

2) Verifies containment isolation valves closure times.

10. PPS 1) Verifies sampling 1) 46.02.01 nonsafety-related capability of the primary 2) 46.02.02 sampling points. 3) 46.02.03

2) Verifies sampling capability of the containment sampling points.

3) Verifies sampling capability of the secondary sampling points.

11. SDIS 1) Verify that the output 1) 59.02.01 nonsafety-related signals from the NPM level, pressure, temperature, and flow instruments listed in Table 7.1-2 properly trend while increasing RCS temperature and pressure.

12. TGS 1) Verifies the TGS 1) 29.02.01 nonsafety-related automatically controls 2) 29.02.01 turbine bypass flow to 3) 29.02.01 the main condenser.

2) Verify the maximum main turbine speed that can be obtained using the MHS to heat the RCS.

3) Verifies the ECCS valves close when the CVCS provides water to reset the ECCS valves.

NuScale US460 SDAA 14.2-140 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-63: Test # 63 Hot Functional Testing (Continued) 63.00.XX Prerequisites Prerequisites associated with performing hot functional testing are identified in the referenced test abstract cited under the Verified by Test # heading.

NuScale US460 SDAA 14.2-141 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-64: Test # 64 Module Assembly Equipment Bolting Preoperational test is required to be performed once.

The MAE 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. The MAE bolting supports the nonsafety-related 45.02.01 CNTS actively by providing material handling to allow for disassembly and reassembly of the containment vessel lower flange.

2. The MAE bolting supports the nonsafety-related 45.02.01 RPV actively by providing material handling to allow for disassembly and reassembly of the RPV lower flange.

64.00.XX Prerequisites

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

64.01.XX Component Level Tests None 64.02.XX System-Level Tests None NuScale US460 SDAA 14.2-142 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-65: Test # 65 Steam Generator Flow-Induced Vibration This is a one-time test to be performed before loading fuel in the first ever NPM. There are no preoperational tests for the SG system.

Validation testing is performed at test facilities as separate effects tests on prototypic SG tubes and functionally equivalent SG tube supports per Section 5.1 of TR-121354-P.

The SG flow-induced vibration testing is performed consistent with the requirements of the NuScale CVAP as described in the NuScale Comprehensive Vibration Assessment Program Analysis Technical Report, TR-121353-P, and the NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report, TR-121354-P. The SG tube testing consists of in-air and in-water modal testing and primary side flow testing. The CVAP is addressed in Section 3.9.2. The SGs are discussed in Section 5.4.1.

System Function System Function Categorization Function Verified by Test #

None N/A N/A 65.00.XX Prerequisites:

N/A 65.01.XX Component Level Tests None Acceptance Criteria:

1) The SG tube testing shows that fluid elastic instability and vortex shedding do not occur under primary side flow rates consistent with any operating condition, considering all applicable uncertainties and biases of this separate effects test.

2) The SG tube testing shows that for primary side flow rates consistent with 100 percent power operation, the SG tube vibration responses are less than those predicted with the turbulent buffeting analysis methodology.

NuScale US460 SDAA 14.2-143 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-66: Test # 66 Security Access Control Preoperational test is required to be performed once.

Security access control is described in NuScale Design of Physical Security Systems, TR-118318.

System Function System Function Categorization Function Verified by Test #

1. The security access controls security-related Component level test 66.01.01 support the security plan described in TR-118318.

66.00.XX Prerequisites

01. Security access control boundary for the protected and vital areas, described in the security technical report, are established.

66.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Verify an access control system 1) Use authorized and unauthorized 1) The access points do not allow with a numbered photo identification badges in all vital access to unauthorized badges.

identification badge system that area access points in the RXB and 2) The access points allow controls access to vital areas CRB identified in NuScale Design authorized personnel.

within the RXB and CRB to of Physical Security Systems," [ITAAC 03.16.04]

authorized personnel. TR-118318.

66.02.XX System Level Tests None NuScale US460 SDAA 14.2-144 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-67: Test # 67 Security Detection and Alarm Preoperational test is required to be performed once.

Security detection and alarm is described in NuScale Design of Physical Security Systems, TR-118318.

System Function System Function Categorization Function Verified by Test #

1. The security detection and alarm security-related Component level tests 67.01.01 -

system acts to satisfy the 67.01.05 functional requirements described in TR-118318.

67.00.XX Prerequisites

01. Required security system SAT is completed and approved.

67.01.XX Component Level Tests Test Objective Test Method Acceptance Criteria

01. Unoccupied vital areas must be 1) Access to all unoccupied vital 1) Verify the access door is locked.

designed with locking devices and areas that are identified in the Upon entry into the room verify an intrusion detection devices that TR-118318. intrusion alarm is received in the annunciate in the central alarm central alarm station.

station. [ITAAC 03.16.05]

02. Security alarm devices including 1) Insert a signal real or simulated 1) Verify alarm annunciation is transmission lines to annunciators tamper signal. received in the central alarm are tamper-indicating and 2) Insert a signal real or simulated of station for each test method. The self-checking. a component failure for all alarm alarm must indicate the type and devices and transmission lines in location of the alarm.

the RXB and CRB. [ITAAC 03.16.07]

3) Place all security alarm devices in the RXB and CRB on standby power.

03. Intrusion detection and 1) Put all intrusion detection 1) Verify an audible and visual alarm assessment systems provides equipment described in TR-118318 is received in the central alarm visual and audible alarm into an alarm state. station.

annunciation in the central alarm [ITAAC 03.16.08]

station.

04. Intrusion detection system 1) Place all intrusion detection 1) Verify the intrusion detection recording equipment records equipment in the RXB and CRB in system recording system records onsite security alarm annunciation the following alarm conditions (as each alarm to include:

including false alarm, alarm applicable to the equipment): a. Location of the alarm check, and tamper indication and a. False alarm b. Type of alarm the type of alarm, location, alarm b. Alarm check c. Alarm circuit circuit, date, and time. c. Tamper indication d. Date

e. Time (this test can be done in conjunction with audible and visual alarm testing)

[ITAAC 03.16.09]

05. Emergency exits in the RXB and 1) Attempt to enter each the RXB and 1) Verify the locking device prevents CRB must be alarmed with CRB exits. entry.

intrusion detection devices and 2) Exit each of the RXB and CRB 2) Verify the exit allows for prompt secured by locking devices that exits. exit of the building and alarms in allow prompt egress during an the central alarm station when emergency. opened.

[ITAAC 03.16.10]

67.02.XX System Level Tests None NuScale US460 SDAA 14.2-145 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-68: Test # 68 Initial Fuel Loading and Precritical The Initial Fuel Loading Precritical Test is required to be performed for each NPM.

This test is performed after initial fuel loading but before initial criticality.

Test Objectives

1. Identify the sequence for precritical testing (after fuel load and before criticality).

2. The precritical tests are:

a. RCS Flow Measurement 70.03.01

b. NPM Temperatures 71.03.01

c. Primary and Secondary System Chemistry 72.03.01

d. CRDS - Manual Operation, Rod Speed, and Rod Position Indication 73.03.01

e. Control Rod Assembly Full-Height Drop Time 74.03.01

f. Control Rod Assembly Ambient Temperature Full-Height Drop Time 75.03.01

g. Pressurizer Spray Bypass Flow 76.03.01 68.00.XX Prerequisites None 68.03.01 Test Method

1. Identify the specific plant conditions required for each precritical test procedure to maintain TS operability.

2. Identify the prerequisites required for each precritical test procedure.

3. Determine the test sequence for precritical testing based on TS requirements and test prerequisites.

Acceptance Criterion

1. The sequence for precritical testing is determined.

NuScale US460 SDAA 14.2-146 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-69: Test # 69 Initial Fuel Load The Initial Fuel Load Test is required to be performed for each NPM.

This test is performed before initial fuel load.

Test Objectives

1. Conduct initial fuel load with no inadvertent criticality.

2. Install fuel assemblies and control components at the locations specified by the design of the initial RXCS.

69.00.XX Prerequisites

01. Plant systems required for initial fuel loading have completed preoperational testing.

02. Plant systems required for initial fuel loading have been aligned per operations procedures.

03. The design of the initial RXCS that specifies the final core configuration of fuel assemblies and control components is completed.

04. A core load sequence is approved.

05. 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.

06. The lower RPV is installed in the RPV support stand.

07. RXB radiation monitors are functional.

08. Boron concentration in the pool is within TS limits.

09. The nuclear instrumentation system is calibrated and operable.

69.03.01 Test Method

1. Install fuel and control components per approved procedures.

2. Monitor boron concentration inside the RPV periodically during fuel load to ensure it satisfies TS.

3. 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.

4. 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.

5. Demonstrate the inverse count rate ratio does not show significant approach to criticality.

6. Maintain the status of the core loading.

Acceptance Criteria

1. Each fuel assembly and control component is installed in the location specified by the design of the initial reactor core.

2. There is no indication of inadvertent criticality.

NuScale US460 SDAA 14.2-147 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-70: Test # 70 Reactor Coolant System Flow Measurement The RCS Flow Measurement Test is required to be performed for each NPM.

This test is performed after initial fuel loading but before initial criticality.

Test Objective

1. Verify that the RCS flow is sufficient to ensure adequate boron mixing in the RCS coolant.

70.00.XX Prerequisites

01. The core is installed.

02. The NPM is fully assembled.

03. The RCS is at HZP (RCS at normal operating pressure with RCS temperature at the maximum temperature obtainable when heated only by the MHS).

04. The RCS flow meters have been calibrated.

70.03.01 Test Method

1. Record RCS flow using MCR indication.

Acceptance Criterion

1. The RCS flow at HZP satisfies the minimum RCS flow assumed in the safety analysis.

NuScale US460 SDAA 14.2-148 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-71: Test # 71 NuScale Power Module Temperatures Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading but before initial criticality.

Test Objectives

1. Perform a cross calibration of the RTDs monitored by the MPS listed in Table 7.1-2.

2. Verify incore thermocouple resistance leakage satisfies manufacturer's criteria.

71.00.XX Prerequisites

01. The core is installed.

02. The NPM is fully assembled.

03. The calibration of reactor coolant system RTDs is completed.

71.03.01 Test Method

1. With the RCS at ambient temperature and isothermal conditions record the following data:

MCR indication of RTD temperatures monitored by MPS

MCR indication of incore thermocouples temperatures

Leakage resistance of the incore thermocouples

2. Increase RCS temperature by approximately 50°F.

3. Record RTD and incore thermocouple data at isothermal conditions.

4. Repeat data collection until RCS temperature is at the highest temperature obtainable using only the module heatup system.

5. Cross-calibrate RTD temperatures monitored by MPS that monitor the same variable.

Acceptance Criteria

1. The cross calibration of the reactor coolant system RTDs is completed.

2. The leakage resistance of the fixed incore detectors satisfies manufacturer's recommendations.

NuScale US460 SDAA 14.2-149 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-72: Test # 72 Primary and Secondary System Chemistry Startup test is required to be performed for each NPM.

This test is performed before criticality and at approximately 25, 50, 75, and 100 percent reactor thermal power.

Test Objective

1. Verify water quality in the primary system and secondary system using the PSS.

72.00.XX Prerequisites

01. The PSS instruments have been calibrated.

02. The NPM is fully assembled.

03. The RCS is at HZP (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).

72.03.01 Test Method

1. Use the PSS to sample the normal primary system sample points listed in Table 9.3.2-1.

2. Use the PSS to sample the normal secondary system sample points listed in Table 9.3.2-3.

3. To the extent practicable, responses of PSS radiation monitors are verified by laboratory analyses of grab samples taken at the same process location.

4. Conduct the test before criticality and at steady-state condition at approximately 25, 50, 75, and 100 percent reactor thermal power.

Acceptance Criterion

1. The sample analyses satisfy the limits specified in plant procedures.

NuScale US460 SDAA 14.2-150 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-73: Test # 73 Control Rod Drive System - Manual Operation, Rod Speed, and Rod Position Indication Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading but before initial criticality.

Test Objectives

1. Verify the ability to manually fully insert and fully withdraw individual CRAs from the MCR.

2. Verify CRA rod position indications provide indication of rod movement.

3. Verify individual CRA position indications are within the required number of steps of their associated group position.

4. Verify the rod insertion and withdrawal speeds are within design limits.

73.00.XX Prerequisites

01. The core is installed.

02. The NPM is fully assembled.

03. The RCS is at HZP (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).

04. All RCS temperatures satisfy the minimum TS temperature for criticality.

05. The nuclear instrumentation system is calibrated and operable.

06. The shutdown margin is within the limits specified in the core operating limits report.

73.03.01 Test Method

1. 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 is withdrawn. All other banks are fully inserted. Repeat the test until all shutdown banks and regulating banks are tested.

2. 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

1. All CRAs can be individually fully withdrawn and fully inserted from the MCR.

2. Individual CRA position indications are within the number of steps of their associated group position as required by TS.

3. The CRA insertion and withdrawal speeds are within the design limits identified in Section 3.9.4.

NuScale US460 SDAA 14.2-151 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-74: Test # 74 Control Rod Assembly Full-Height Drop Time Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading but before initial criticality.

Test Objective

1. Verify each CRA satisfies the CRA drop time acceptance criteria for RCS flow at 0 percent reactor thermal power.

74.00.XX Prerequisites

01. The core is installed.

02. The NPM is fully assembled.

03. The RCS is at HZP (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).

04. All RCS temperatures satisfy the minimum TS temperature for criticality.

05. The nuclear instrumentation system is calibrated and operable.

06. The shutdown margin is within the limits specified in the core operating limits report.

74.03.01 Test Method

1. Fully withdraw each individual CRA.

2. Interrupt the electrical power to the associated CRDM.

3. Measure the CRA drop time.

Acceptance Criteria

1. Each CRA drop time is within TS limits.

2. Each CRA drop time is within two sigma of the drop time data for all control rods, or is verified within TS limits by a minimum of three additional performances of this test.

NuScale US460 SDAA 14.2-152 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-75: Test # 75 Control Rod Assembly Ambient Temperature Full-Height Drop Time Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading but before initial criticality.

Test Objective

1. Verify each CRA satisfies the CRA drop time acceptance criteria for RCS at ambient temperature.

75.00.XX Prerequisites

01. The core is installed.

02. The NPM is fully assembled.

03. The RCS is at cold temperature conditions.

04. The nuclear instrumentation system is calibrated and operable.

05. The shutdown margin is within the limits specified in the core operating limits report.

75.03.01 Test Method

1. Fully withdraw each individual CRA.

2. Interrupt the electrical power to the associated CRDM.

3. Measure the CRA drop time.

Acceptance Criteria

1. Each CRA drop time is within TS limits.

2. Each CRA drop time is within two sigma of the drop time data for all control rods, or is verified within TS limits by a minimum of three additional performances of this test.

NuScale US460 SDAA 14.2-153 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-76: Test # 76 Pressurizer Spray Bypass Flow Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading but before initial criticality.

Test Objective

1. Verify the PZR spray bypass flow rate is adequate to prevent thermal fatigue of the spray line components and provide sufficient mixing in the PZR to maintain PZR water chemistry similar to the rest of the RCS while avoiding unnecessary energization of the PZR heaters.

76.00.XX Prerequisites

01. The core is installed.

02. The NPM is fully assembled.

03. The RCS is at HZP (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).

76.03.01 Test Method

1. With the automatic PZR spray valve closed, adjust the manual spray bypass valve to maintain a continuous spray bypass flow of approximately one gpm.

2. If the continuous bypass spray flow requires the operation of the PZR backup heaters to maintain the PZR pressure setpoint, throttle close the bypass valve until PZR pressure is maintained by the proportional heaters.

Acceptance Criterion

1. The spray bypass valve flow satisfies design requirements.

NuScale US460 SDAA 14.2-154 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-77: Test # 77 Initial Criticality Startup test is required to be performed for each NPM.

This test is performed after initial fuel loading.

Test Objective

1. Achieve initial criticality in a controlled manner.

77.00.XX Prerequisites

01. The RCS is at HZP (RCS at normal operating pressure and RCS temperature at the maximum temperature obtainable when heated only by the MHS).

02. All RCS temperatures satisfy the minimum TS temperature for criticality.

03. The nuclear instrumentation system is calibrated and operable.

04. The shutdown margin is within the limits specified in the core operating limits report.

05. An estimated critical position (calculation) is performed.

06. RCS measured boron is at or near the desired estimated critical position value.

07. The shutdown banks and the regulating banks are fully inserted.

08. A neutron count rate of at least 1/2 counts per second registers on the startup channels, and the signal to noise ratio is greater than 2.

77.03.01 Test Method

1. Shutdown banks are withdrawn in sequence using the sequence of a normal plant startup. Gather data to plot the inverse.

2. Count rate ratio. The inverse count rate ratio is used to monitor reactivity.

3. 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.

4. After criticality is obtained, the regulating bank is confirmed to be above the TS regulating group insertion limit.

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 are increased until the regulating bank is withdrawn sufficiently to meet the insertion limit.

Acceptance Criterion

1. The reactor is critical with the regulating banks above their TS insertion limit.

NuScale US460 SDAA 14.2-155 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-78: Test # 78 Post-Critical Reactivity Computer Checkout Startup test is required to be performed for each NPM.

This test is performed after initial criticality.

Test Objective

1. Verify proper operation of the reactivity computer to measure reactivity changes in the core during low-power testing.

78.00.XX Prerequisites

01. The reactor is critical with the neutron flux level within the range for low-power physics testing.

02. The RCS temperature and pressure are stable at the normal no-load values.

03. The neutron flux level and RCS boron concentration are stable.

04. The reactivity computer is installed and internal reactivity computer checks have been completed.

78.03.01 Test Method

1. Withdraw the regulating bank to achieve a positive startup rate below TS limits.

2. Measure the reactor period or doubling time.

3. Reinsert the regulating bank to re-establish the initial steady-state neutron flux.

4. Measure the negative reactor period or halving time.

5. Validate the core response against the reactivity computer input delayed neutron fractions and prompt neutron lifetime using pre-determined test criteria.

6. Adjust and recalibrate reactivity computer until acceptance criteria are met.

Acceptance Criterion

1. The reactivity computer is calibrated.

NuScale US460 SDAA 14.2-156 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-79: Test # 79 Low-Power Test Sequence Startup test is required to be performed for each NPM.

This test is performed before initial criticality.

Test Objectives

1. Identify the sequence for low-power testing.

2. The low-power tests are:

a. Determination of Zero-Power Physics Testing Range 80.03.01

b. All Rods Out Boron Endpoint Determination 81.03.01

c. Isothermal Temperature Coefficient Measurement 82.03.01

d. Bank Worth Measurement 83.03.01 79.00.XX Prerequisites None 79.03.01 Test Method For each of the tests identified in the test objectives above:

1. Identify the specific plant conditions required for each low-power test procedure to maintain TS operability.

2. Identify the prerequisites required for each low-power test procedure.

3. Determine the test sequence for low-power testing based on TS requirements and test prerequisites.

Acceptance Criterion

1. The sequence for low-power testing is determined.

NuScale US460 SDAA 14.2-157 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-80: Test # 80 Determination of Zero-Power Physics Testing Range Startup test is required to be performed for each NPM.

This test is performed after initial criticality.

Test Objectives

1. Determine the reactor flux level at which the point of nuclear heating is detectable.

2. Establish the range of neutron flux in which HZP reactivity measurements are to be performed.

80.00.XX Prerequisites

01. The reactor is critical with the neutron flux level at steady-state below the expected level of nuclear heating.

02. The RCS temperature and pressure is steady-state at the normal HZP conditions.

03. The RCS boron concentration is steady-state.

04. The reactivity computer is operational and recording the core average neutron flux level.

05. The regulating bank is positioned to allow reactivity changes by rod motion alone.

80.03.01 Test Method

1. Withdraw the regulating bank to establish a slow startup rate allowing neutron flux level to increase until nuclear heating is observed.

2. Record the reactivity computer neutron flux level and the corresponding MCR flux indication at which nuclear heating occurs.

3. Insert the regulating bank to establish a reactivity computer flux level about one-third of the value at which nuclear heating is observed. This flux level becomes the maximum value for the zero-power testing range.

Acceptance Criterion

1. The zero-power testing range flux level is determined.

NuScale US460 SDAA 14.2-158 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-81: Test # 81 All Rods Out Boron Endpoint Determination Startup test is required to be performed for each NPM.

This test is performed after initial criticality.

Test Objective

1. Determine the critical RCS boron concentration for all rods out (ARO) (fully withdrawn shutdown banks and regulating banks) at HZP.

81.00.XX Prerequisites

01. The reactor is critical with the neutron flux level at steady-state below the expected level of nuclear heating.

02. The RCS temperature and pressure is steady-state at the normal HZP conditions.

03. The RCS boron concentration is steady-state.

04. The reactivity computer is operational and recording the core average neutron flux level.

81.03.01 Test Method

1. 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 is near fully withdrawn and limits the usable positive reactivity remaining in the rods with the reactor critical.

2. Measure the just-critical boron concentration by chemical analysis.

3. 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

1. The measured value for the ARO boron endpoint satisfies the design value contained within the test acceptance criteria.

NuScale US460 SDAA 14.2-159 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-82: Test # 82 Isothermal Temperature Coefficient Measurement Startup test is required to be performed for each NPM.

This test is performed after initial criticality.

Test Objectives

1. Determine the isothermal temperature coefficient.

2. Calculate the moderator temperature coefficient.

82.00.XX Prerequisites

01. The reactor is critical with the neutron flux level at steady-state below the expected level of nuclear heating.

02. The RCS temperature and pressure is steady-state at the normal HZP conditions.

03. The RCS boron concentration is steady-state.

04. The reactivity computer is operational and recording the core average neutron flux level.

05. The regulating rod bank is positioned near fully withdrawn (near their ARO position).

82.03.01 Test Method

1. Vary RCS temperature (heatup/cooldown) while maintaining rods and boron concentration constant.

2. Monitor reactivity results and determine the isothermal temperature coefficient.

3. Calculate the moderator temperature coefficient using the isothermal temperature coefficient and design values.

Acceptance Criterion

1. The moderator temperature coefficient is within the limits specified in the core operating limits report.

NuScale US460 SDAA 14.2-160 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-83: Test # 83 Bank Worth Measurement Startup test is required to be performed for each NPM.

This test is performed after initial criticality.

Test Objectives

1. Measure the integral and differential worth of the reference bank (the test bank with the highest predicted worth).

2. Measure the worth of the remaining shutdown and regulating banks by control rod exchange (rod swap).

83.00.XX Prerequisites

01. The reactor is critical with the neutron flux level at steady-state within the range for HZP physics testing.

02. The RCS temperature and pressure is steady-state at the normal HZP conditions.

03. The RCS boron concentration is steady-state.

04. The reactivity computer is operational and recording the core average neutron flux level.

05. The regulating rod banks are positioned near fully withdrawn (near their ARO position).

83.03.01 Test Method

1. 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.

2. 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

1. The measured worth for each individual bank, and sum of bank worths, is consistent with the predicted value within the test acceptance criteria.

NuScale US460 SDAA 14.2-161 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-84: Test # 84 Power-Ascension Startup test is required to be performed for each NPM.

This test is performed before power-ascension testing.

Test Objective Identify the sequence for the following power-ascension tests.

a. Core Power Distribution Map 85.03.01

b. Neutron Monitoring System Power Range Flux Calibration 86.03.01

c. RCS Temperature Instrument Calibration 87.03.01

d. RCS Flow Calibration 88.03.01

e. Radiation Shield Survey 89.03.01

f. RBVS Capability 90.03.01

g. Thermal Expansion 91.03.01

h. Control Rod Assembly Misalignment 92.03.01

i. SG Level Control System 93.03.01

j. Ramp Change in Load Demand 94.03.01

k. Step Change in Load Demand 95.03.01

l. Loss of FWH 96.03.01

m. 100 Percent Load Rejection 97.03.01

n. Reactor Trip from 100 Percent Power 98.03.01

o. Island Mode Test for the First NPM 99.03.01

p. Island Mode Test for Multiple NPMs 100.03.01

q. NPM Vibration 102.03.01 84.00.XX Prerequisites None 84.03.01 Test Method

1. Identify the specific plant conditions required for each power-ascension test procedure to maintain TS operability.

2. Identify the prerequisites required for each power-ascension test procedure.

3. Determine the test sequence for power-ascension testing based on TS requirements and test prerequisites.

Acceptance Criterion

1. The sequence for power-ascension testing is determined.

NuScale US460 SDAA 14.2-162 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-85: Test # 85 Core Power Distribution Map 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

1. Obtain a core power distribution map during power ascension.

2. Using the data from the core power distribution map verify core power distribution is consistent with design predictions and associated TS limits.

85.00.XX Prerequisites

01. The ICIS is operational.

02. The NPM is operating in a steady-state condition at the specified power level.

03. Maintain reactor power, TAVG, and PZR level constant during data collection.

85.03.01 Test Method

1. 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.

2. Use data from the in-core maps to verify that core power distribution is consistent with design predictions and TS limits.

Acceptance Criterion

1. Core power distribution is consistent with design predictions and TS limits.

NuScale US460 SDAA 14.2-163 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-86: Test # 86 Neutron Monitoring System Power Range Flux Calibration 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

1. Calibrate the NMS power range neutron flux signals during power ascension.

86.00.XX Prerequisites

01. The ICIS is operational.

02. The NPM is operating in a steady-state condition at the specified power level.

86.03.01 Test Method

1. 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

heat balance data

2. Maintain reactor power, TAVG, and PZR level constant during data collection.

3. Calibrate the NMS neutron flux power range (linear power) signal using the recorded data.

Acceptance Criterion

1. The NMS neutron flux power range (linear power) signal is calibrated.

NuScale US460 SDAA 14.2-164 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-87: Test # 87 Reactor Coolant System Temperature Instrument Calibration 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

1. Calibrate narrow range RCS hot leg temperature instruments, wide range RCS hot leg temperature instruments, and narrow range RCS cold leg temperature instruments.

87.00.XX Prerequisites

01. The ICIS is operational.

02. The NPM is operating in a steady-state condition at the specified power level.

87.03.01 Test Method

1. 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

2. Maintain reactor power, TAVG, and PZR level at steady-state during data collection.

3. 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

1. The RCS hot and cold let temperature instruments have been calibrated.

NuScale US460 SDAA 14.2-165 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-88: Test # 88 Reactor Coolant System Flow Calibration 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

1. Calibrate the RCS flow instruments during power ascension.

88.00.XX Prerequisites

01. The ICIS is operational.

02. The NPM is operating in a steady-state condition at the specified power level.

03. The nuclear instrumentation system is calibrated and operable.

88.03.01 Test Method

1. 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

2. Maintain reactor power, TAVG, and PZR level at steady state during data collection.

3. Calibrate the RCS flow instruments using the recorded data.

Acceptance Criterion

1. The RCS flow instruments have been calibrated.

NuScale US460 SDAA 14.2-166 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-89: Test # 89 Radiation Shield Survey 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

1. Verify the adequacy of radiation shields in the RXB designed to protect personnel from radiation originating from sources within the reactor vessel.

89.00.XX Prerequisites

01. Radiation survey instruments are calibrated.

02. The NPM is operating in a steady-state condition at the specified power level.

89.03.01 Test Method

1. Measure gamma and neutron radiation dose rates at designated locations at approximately 25, 50, and 100 percent reactor thermal power in accordance with RG 1.69 and ANSI/ANS-6.3.1 (1987, R2007).

2. The designated locations are the accessible areas outside permanent radiation shields in the RXB.

Acceptance Criterion

1. Radiation dose rates are consistent with design expectations.

NuScale US460 SDAA 14.2-167 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-90: Test # 90 Reactor Building Ventilation System Capability 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

1. Verify that the RBVS maintains the design environment in areas containing equipment that is environmentally qualified for a harsh or mild environment.

90.00.XX Prerequisite

01. The NPM is operating in a steady-state condition at the specified power level.

90.03.01 Test Method

1. With the plant at power levels of approximately 50 and 100 percent of reactor thermal power and RBVS in normal lineup, record temperature and humidity for the environmental qualification zones listed in Table 3C-1 that are not under the bioshield.

2. With the plant at power levels of approximately 50 and 100 percent of reactor thermal power and RBVS in normal lineup, record the temperature and humidity in the rooms containing electrical equipment qualified for a mild environment.

Acceptance Criteria

1. Room temperature and humidity in environmental qualification zones listed in Table 3C-1 that are not under the bioshield satisfy the indoor design conditions for the RBVS contained in Table 9.4.2-1.

2. Room temperature and humidity in rooms containing electrical equipment qualified for a mild environment satisfy the indoor design conditions for the RBVS contained in Table 9.4.2-1.

NuScale US460 SDAA 14.2-168 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-91: Test # 91 Thermal Expansion Startup test is required to be performed for each NPM.

This test is performed during plant heatup and cooldown.

Test Objectives

1. 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.

2. Verify that high-energy piping inside the RXB can expand without obstruction and that expansion is within design limits.

91.00.XX Prerequisite

01. Temporary instrumentation is installed on piping as required to monitor the deflections for the piping under test.

91.03.01 Test Method

1. Thermal expansion testing is performed in accordance with ASME OM Code, Division 3, Part 7 as discussed in Section 3.9.2.1.

2. Record deflection data during plant heatup and cooldown.

3. Identify support movements by recording hot and cold positions of the supports.

Acceptance Criteria In accordance with ASME OM Code, Division 3, Part 7, for the piping systems tested:

1. There is no evidence of constrained thermal expansion of piping or components, other than by installed supports and restraints that are designed to prevent thermal movement.

2. Pipe support movements must be within manufacturer specifications.

3. Piping and components return to their approximate baseline cold position.

NuScale US460 SDAA 14.2-169 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-92: Test # 92 Control Rod Assembly Misalignment Startup test is required to be performed for each NPM.

This test is performed at approximately 50 and 100 percent reactor thermal power.

Test Objectives

1. Verify that core thermal and nuclear parameters at 50 and 100 percent 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.

2. Verify the capability of the in-core neutron flux instrumentation to detect a control rod misalignment equal to or less than the TS limits at 50 and 100 percent reactor thermal power.

3. Monitor the power distribution following the recovery of a misaligned CRA.

92.00.XX Prerequisites

01. The reactor is operating at steady-state conditions and is at that condition for a sufficient time to reach xenon equilibrium.

02. The reactor power level, RCS boron concentration, and temperature are stable.

03. 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.

92.03.01 Test Method

1. 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.

2. 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.

3. Record incore and excore instrumentation signals to determine their response and to determine the power distribution and power peaking factors before CRA misalignment, at partial misalignment, at full misalignment, and periodically after restoration to normal.

Acceptance Criteria

1. Measured power distributions and power peaking factors are within TS limits and are consistent with the predictions.

NuScale US460 SDAA 14.2-170 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-93: Test # 93 Steam Generator Level Control 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

1. Verify the ability of SG inventory control systems to sustain a ramp increase in load demand.

2. Assess the dynamic response of SG inventory for ramp increase in load demand.

93.00.XX Prerequisite

01. The FWS is operating in SG inventory pressure control (FW regulating valves in automatic control).

93.03.01 Test Method

1. Raise reactor thermal power to approximately 25 percent.

2. Use the MCR turbine controls to provide a 5 percent of full power per minute load increase in demand at approximately 25, 50, and 75 percent reactor thermal power.

3. Use the MCR turbine controls to provide a 5 percent of full power per minute load decrease in demand at approximately 25, 50, and 75, and 100 percent reactor thermal power.

Acceptance Criteria

1. The SG inventory control systems, with no manual intervention, maintain the following parameters within design limits during and following the transient:

a. SG superheat

b. SG pressure

c. SG inventory

d. Feed pump speed

2. The SG inventory control systems response is reviewed and compared to expected performance. Necessary adjustments to the control systems have been made before proceeding to the next power plateau.

NuScale US460 SDAA 14.2-171 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-94: Test # 94 Ramp Change in Load Demand 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

1. Verify the ability of the plant automatic control systems to sustain a ramp increase in load demand.

a. Assess the dynamic response of the plant for ramp increase in load demand.

94.00.XX Prerequisites

01. The NPM is operating in a steady-state condition at the designated power level.

02. The plant's electrical distribution system is aligned for normal operation.

03. The following control systems are in automatic control:

a. Reactivity control

b. RCS temperature control

c. PZR pressure control

d. PZR level control

e. Turbine control

f. FW level control

g. SCWS basin level control

h. FWH level control

i. CCT level control

04. If required, verify instrumentation is installed for piping vibration testing.

94.03.01 Test Method

1. Use the MCR turbine controls to provide a 5 percent of full power per minute load increase in demand at approximately 25, 50, and 75 percent reactor thermal power.

2. Use the MCR turbine controls to provide a 5 percent of full power per minute load decrease in demand at approximately 25, 50, and 75, and 100 percent reactor thermal power.

3. Conduct piping vibration testing, as required, during power changes.

NuScale US460 SDAA 14.2-172 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-94: Test # 94 Ramp Change in Load Demand (Continued)

Acceptance Criteria

1. The turbine does not trip.

2. The reactor does not trip.

3. The MS safety valves do not open.

4. The turbine does not overspeed.

5. The plant automatic control systems, with no manual intervention, maintain the following parameters within design limits during and following the transient:

a. Reactor power

b. RCS temperature

c. PZR pressure

d. PZR level

e. SG superheat

f. SG pressure

g. SG inventory

h. Gland seal temperature

i. SCWS basin level

j. FWH level

k. CCT level

l. Main condenser vacuum

m. Outlet temperature of turbine bypass desuperheater

6. Control system response is reviewed and compared to expected performance. Necessary adjustments to the control systems have been made before proceeding to the next power plateau.

7. 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.

8. Piping vibration - System specific steady state vibration testing criteria are established by the piping designer.

Actual acceptance criteria depends on the selected test method, but may include:

a. Limits for stresses calculated based on the observed/measured vibration response of the system.

b. No permanent deformation or damage is observed in the piping system or supports.

c. Vibration displacements are not excessive, may not potentially cause the piping to come in contact with surrounding SSC, and are such that the movement of supports and flexible joints is within their allowable limits.

NuScale US460 SDAA 14.2-173 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-95: Test # 95 Step Change in Load Demand 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

1. Verify the ability of the plant automatic control systems to sustain step load increases and step load decreases in demand.

2. Assess the dynamic response of the plant for a load step demand.

95.00.XX Prerequisites

01. The NPM is operating in a steady-state condition at the specified power level.

02. The plant's electrical distribution system is aligned for normal operation.

03. The following control systems are in automatic control:

a. Reactivity control

b. RCS temperature control

c. PZR pressure control

d. PZR level control

e. Turbine control

f. FW level control

g. SCWS basin level control

h. FWH level control

i. CCT level control 95.03.01 Test Method

1. Use the MCR turbine controls to provide a 10 percent step load increase in demand at approximately 25, 50, and 75 percent reactor thermal power.

2. Use the MCR turbine controls to provide a 10 percent step load decrease in demand at approximately 25, 50, 75, and 100 percent reactor thermal power.

Acceptance Criteria

1. The turbine does not trip.

2. The reactor does not trip.

3. The MS safety valves do not open.

4. The turbine does not overspeed.

5. The plant automatic control systems, with no manual intervention, maintain the following parameters within design limits during and following the transient:

a. Reactor power

b. RCS temperature

c. PZR pressure

d. PZR level

e. SG superheat

f. SG pressure

g. SG inventory

h. Gland seal temperature

i. SCWS basin level

j. FWH level

k. CCT level

l. Main condenser vacuum

m. Outlet temperature of turbine bypass desuperheater

6. Control system response is reviewed and compared to expected performance. Necessary adjustments to the control systems have been made before proceeding to the next power plateau.

7. 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.

NuScale US460 SDAA 14.2-174 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-96: Test # 96 Loss of Feedwater Heater 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

1. Verify the ability of the plant automatic control systems to sustain a loss of the high pressure FWH during power operation.

2. Assess the dynamic response of the plant for the loss of the high pressure FWH.

96.00.XX Prerequisites

01. The NPM is operating in a steady-state condition at the specified power level.

02. The plant's electrical distribution system is aligned for normal operation.

03. The following control systems are in automatic control:

a. Reactivity control

b. RCS temperature control

c. PZR pressure control

d. PZR level control

e. Turbine control

f. FW level control

g. SCWS basin level control

h. FWH level control

i. CCT level control 96.03.01 Test Method

1. Close the turbine generator extraction steam supply isolation valve to the high pressure FWH from the MCR at approximately 50 and 90 percent reactor thermal power.

Acceptance Criteria

1. The reactor does not trip.

2. The turbine does not trip.

3. The MS safety valves do not open.

4. The plant automatic control systems, with no manual intervention, maintain the following parameters within design limits during and following the transient:

a. Reactor power

b. RCS temperature

c. PZR pressure

d. PZR level

e. SG superheat

f. SG pressure

g. SG inventory

h. Gland seal temperature

i. SCWS basin level

j. FWH level

k. CCT level

l. Main condenser vacuum

m. Outlet temperature of turbine bypass desuperheater NuScale US460 SDAA 14.2-175 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-97: Test # 97 100 Percent Load Rejection Startup test is required to be performed for each NPM.

This test is performed at approximately 100 percent reactor thermal power.

Test Objectives

1. Verify the ability of the plant automatic control systems to sustain a 100 percent load rejection from full power.

2. Assess the dynamic response of the plant for a 100 percent power load rejection.

97.00.XX Prerequisites

01. The NPM is operating in a steady-state condition at full reactor thermal power.

02. The plant's electrical distribution system is aligned for normal operation.

03. The following control systems are in automatic control:

a. Reactivity control

b. RCS temperature control

c. PZR pressure control

d. PZR level control

e. Turbine control

f. FW level control

g. SCWS basin level control

h. FWH level control

i. CCT level control 97.03.01 Test Method

1. Manually trip the generator output breaker to provide a 100 percent load rejection.

Acceptance Criteria

1. The turbine trips.

2. The reactor does not trip.

3. The MS safety valves do not open.

4. The turbine does not overspeed beyond design limits.

5. The turbine generator bypass valve opens and modulates steam flow to the condenser to maintain steam generator pressure.

6. The plant automatic control systems, with no manual intervention, maintain the following parameters within design limits during and following the transient:

a. Reactor power

b. RCS temperature

c. PZR pressure

d. PZR level

e. SG superheat

f. SG inventory

g. Gland seal temperature

h. SCWS basin level

i. FWH level

j. CCT hotwell level

k. Main condenser vacuum

l. Outlet temperature of turbine bypass desuperheater

7. 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.

NuScale US460 SDAA 14.2-176 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-98: Test # 98 Reactor Trip from 100 Percent Power Startup test is required to be performed for each NPM.

This test is performed at 100 percent reactor thermal power.

Test Objectives

1. Assess the dynamic response of the plant to a reactor trip.

2. Verify each fully withdrawn CRA satisfies the CRA drop time acceptance criteria at full flow conditions.

3. Verify the ability of the DHRS to cool the RCS to Mode 3 (all RCS temperatures < 345 °F).

98.00.XX Prerequisites

01. The NPM is operating in a steady-state condition at full reactor thermal power.

02. The plant's electrical distribution system is aligned for normal operation.

98.03.01 Test Method

1. Manually trip the reactor from the MCR.

2. Measure the drop time for each fully withdrawn CRA.

3. Allow the RCS temperature trends to stabilize.

4. Manually initiate DHRS.

5. Allow the RCS to cool to mode 3 after DHRS actuation.

Acceptance Criteria Acceptance criteria to be verified after manual reactor trip:

1. The reactor trips.

2. The turbine generator bypass valve operates to prevent opening of the MS safety valve.

3. The turbine trips.

4. 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.

Acceptance criteria to be verified after DHRS actuation:

5.

a. DHRS actuation valves open.

b. MSIVs close.

c. FWIVs close.

d. FW regulating valves close

e. Secondary MSIVs close

f. Secondary MSIBVs close

g. PZR heater breakers trip

6. The RCS cools to a stable condition in mode 3 (all RCS temperatures < 345 °F) without operator intervention.

7. The RCS cooldown rate is within TS limits.

8. Each fully withdrawn CRA drop time is within TS limits.

NuScale US460 SDAA 14.2-177 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-99: Test # 99 Island Mode Test for the First NuScale Power Module This startup test is required to be performed for the first NPM in power operation. No other NPMs are in power operation. This test is performed once per facility. Startup Test 100.03.01 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.

Test Objective for the first NPM in power operation

1. Verify the first NPM in power operation can operate independently from an offsite transmission grid after transition from the transmission grid to island mode.

2. 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.

99.00.XX Prerequisites

01. The first NPM in power operation is in normal operation at 100 percent reactor thermal power.

99.03.01 Test Method

1. Simulate a loss of the transmission grid by opening the switchyard supply breakers.

Acceptance Criteria 1.

a. The service unit generator 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.

2. The plant electrical loads are transitioned back to the external offsite grid connection when it becomes available.

NuScale US460 SDAA 14.2-178 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-100: Test # 100 Island Mode Test for Multiple NuScale Power Modules This startup test is required to be performed once with multiple (at least two) NPMs in operation. This test is performed once per facility. Startup Test 99.03.01 tests island mode for a single NPM.

Island mode operation is described in Section 8.3.1.

Test Objective for multiple NPMs in operation:

1. Verify all NPMs under test can operate independently from an offsite transmission grid after transition from the transmission grid to island mode.

2. 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.

100.00.XX Prerequisites

01. The NPMs selected for test are in normal operation at 100 percent reactor thermal power.

100.03.01 Test Method

1. Simulate a loss of the transmission grid by opening the switchyard supply breakers.

Acceptance Criteria 1.

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.

2. The plant electrical loads are successfully transitioned back to an external offsite grid connection when it becomes available.

NuScale US460 SDAA 14.2-179 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-101: Test # 101 Remote Shutdown Controls and Monitoring Remote shutdown controls and monitoring is described in Section 7.1.1. Testing associated with remote shutdown controls and monitoring occurs during the performance of FAT and SAT as described below.

Remote shutdown controls and monitoring provides additional, identical, sets of MCS and PCS operator workstations at alternate locations to monitor the NPM status and operate the MCS and PCS during an MCR evacuation. The ability to activate the nonsafety MCS and PCS displays and controls at these alternate locations are verified during SAT. The ability to isolate the safety-related MCR module protection system manual switches using the MCR isolation switches located outside the control room as described in Section 7.2.12 are verified during module protection system FAT and SAT.

Table 14.2-54: Module Control System Test # 54 and Table 14.2-55: Plant Control System Test # 55 contain details regarding MCS and plant control system FAT and SAT.

NuScale US460 SDAA 14.2-180 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-102: Test # 102 NuScale Power Module Vibration This startup test is required to be performed once for the first NPM to be tested. This test supports FOAK testing described in Section 14.2.3.3.

This test is performed during the load ramp from zero to 100 percent power and at 100 percent reactor thermal power. The NPM vibration testing is described in Section 3.9.2; and NuScale Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report, TR-121354-P. This test is coordinated with Test 94.03.01 and Test 98.03.01.

Test Objective for the first NPM to be tested

1. Perform vibration testing of CNTS main steam line branch connections, including DHRS steam piping, and MS drain valve branches during the load ramp up to and at 100 percent reactor thermal power to verify vibration amplitudes in the piping regions confirm there is no acoustic resonance (AR) response.

2. Perform vibration monitoring of the NPM using the signals from dynamic pressure sensors. More details regarding the instrumentation locations and vibration mechanisms being monitored are provided in Section 6.0 of TR-121354-P. Vibration monitoring must be performed during the load ramp up to and at 100 percent reactor thermal power and during a test of DHRS actuation, which are coordinated with Test 98.03.01.

102.00.XX Prerequisites

01. The DHRS steam piping and MS drain valve branches are instrumented to obtain AR data.

02. The NPM is instrumented in accordance with Section 6.0 of TR-121354-P to provide vibration monitoring.

102.03.01 Test Method

1. Perform load ramp up to 100 percent power, then operate the NPM for a sufficient duration at 100 percent power to ensure one million vibration cycles for the component with the lowest structural natural frequency.

2. Monitor the vibration of the CNTS steam piping branches, including the DHRS steam lines and MS drain valve branches. Also monitor the signals of the dynamic pressure sensors. If an unacceptable vibration response develops at any time during initial startup testing, the test conditions must be adjusted to stop the vibration and the reason for the vibration anomaly are investigated before continuing with the testing.

Acceptance Criteria

1. Measured vibration amplitudes in the CNTS steam piping branches confirm there is no acoustic resonance concern.

2. Measured vibration responses in the NPM confirm there are no resonant peaks that could indicate a strongly-coupled flow induced vibration mechanism.

NuScale US460 SDAA 14.2-181 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-103: List of Test Abstracts Test Number System Test Abstract Abbreviation 01 PCWS Pool Cooling and Cleanup System 02 UHS Ultimate Heat Sink 03 PLDS Pool Leakage Detection System 04 RCCWS Reactor Component Cooling Water System 05 CHWS Chilled Water System 06 ABS Auxiliary Boiler System 07 ACC Air Cooled Condenser 08 SCWS Site Cooling Water System 09 PWS Potable Water System 10 UWS Utility Water System 11 DWS Demineralized Water System 12 NDS Nitrogen Distribution System 13 SAS Service Air System 14 IAS Instrument and Control Air System 15 CRHS Control Room Habitability System 16 CRVS Normal Control Room HVAC System 17 RBVS Reactor Building HVAC System 18 RWBVS Radioactive Waste Building HVAC System 19 TBVS Turbine Building HVAC System 20 RWDS Radioactive Waste Drain System 21 BPDS Balance-of-Plant Drain System 22 FPS Fire Protection System 23 FDS Fire Detection System 24 MSS Main Steam System 25 FWS Condensate and Feedwater System 26 FWTS Feedwater Treatment System 27 CPS Condensate Polishing System 28 HVDS Feedwater Heater Vents and Drains System 29 TGS Turbine Generator System 30 LRWS Liquid Radioactive Waste System 31 GRWS Gaseous Radioactive Waste System 32 SRWS Solid Radioactive Waste System 33 CVCS Chemical and Volume Control System 34 BAS Boron Addition System 35 MHS Module Heatup System 36 CES Containment Evacuation System 37 CFDS Containment Flooding and Drain System 38 CNTS Containment System 39 RCS Reactor Coolant System 40 ECCS Emergency Core Cooling System 41 DHRS Decay Heat Removal System 42 ICIS In-core Instrumentation System 43 MAE Module Assembly Equipment 44 FHE Fuel Handling Equipment 45 RBC Reactor Building Crane 46 PSS Process Sampling System 47 EHVS High Voltage AC Electrical Distribution System 48 EMVS Medium Voltage AC Electrical Distribution System 49 ELVS Low Voltage AC Electrical Distribution System 50 EDAS Augmented DC Power System NuScale US460 SDAA 14.2-182 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-103: List of Test Abstracts (Continued)

Test Number System Test Abstract Abbreviation 51 EDNS Normal DC Power System 52 BPSS Backup Power Supply System 53 PLS Plant Lighting System 54 MCS Module Control System 55 PCS Plant Control System 56 MPS Module Protection System 57 PPS Plant Protection System 58 NMS Neutron Monitoring System 59 SDIS Safety Display and Indication System 60 RMS Fixed-Area Radiation Monitoring System 61 COMS Communication System 62 SMS Seismic Monitoring System 63 HFT Hot Functional Testing 64 MAEB Module Assembly Equipment Bolting 65 SG Steam Generator Flow-Induced Vibration 66 N/A Security Access Control 67 N/A Security Detection and Alarm 68 N/A Initial Fuel Loading and Precritical 69 N/A Initial Fuel Load 70 N/A Reactor Coolant System Flow Measurement 71 N/A NuScale Power Module Temperatures 72 N/A Primary and Secondary System Chemistry 73 N/A Control Rod Drive System-Manual Operation, Rod Speed, and Rod Position Indication 74 N/A Control Rod Assembly Full-Height Drop Time 75 N/A Control Rod Assembly Ambient Temperature Full-Height Drop Time Test 76 N/A Pressurizer Spray Bypass Flow 77 N/A Initial Criticality 78 N/A Post-Critical Reactivity Computer Checkout 79 N/A Low-Power Test Sequence 80 N/A Determination of Zero-Power Physics Testing Range 81 N/A All Rods Out Boron Endpoint Determination 82 N/A Isothermal Temperature Coefficient Measurement 83 N/A Bank Worth Measurement 84 N/A Power-Ascension 85 N/A Core Power Distribution Map 86 N/A Nuclear Monitoring System Power Range Flux Calibration 87 N/A Reactor Coolant System Temperature Instrument Calibration 88 N/A Reactor Coolant System Flow Calibration 89 N/A Radiation Shield Survey 90 N/A Reactor Building Ventilation System Capability 91 N/A Thermal Expansion 92 N/A Control Rod Assembly Misalignment 93 N/A Steam Generator Level Control 94 N/A Ramp Change in Load Demand 95 N/A Step Change in Load Demand 96 N/A Loss of Feedwater Heater 97 N/A 100 Percent Load Rejection 98 N/A Reactor Trip from 100 Percent Power 99 N/A Island Mode Test for the First NuScale Power Module NuScale US460 SDAA 14.2-183 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-103: List of Test Abstracts (Continued)

Test Number System Test Abstract Abbreviation 100 N/A Island Mode Test for Multiple NuScale Power Modules 101 N/A Remote Shutdown Workstation 102 N/A NuScale Power Module Vibration NuScale US460 SDAA 14.2-184 Revision 0

NuScale Final Safety Analysis Report Initial Plant Test Program Table 14.2-104: Initial Test Program 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 #38.02.01

valve response to manual ESF #56.02.04 actuation at hot functional test pressure and temperature

valve response time test at hot #56.02.05 functional test pressure and temperature

valve response to manual reactor #97 trip at 100% power ECCS valve design

valve response to manual ESF #56.02.04 actuation at hot functional test pressure and temperature

test of valve IAB at design pressure #56.02.04 ECCS operation

Containment response to ECCS #40.02.01 operation at hot functional test pressure and temperature.

DHRS valve design

valve response to manual ESF #56.02.04 actuation at hot functional test pressure and temperature

valve response to manual reactor #97 trip at 100% power DHRS heat exchanger design

heat exchanger response to manual #41.02.01 ESF actuation at hot functional test pressure and temperature

heat exchanger response to manual #97 reactor trip at 100% power CFDS

automatic fill of containment #37.02.02

automatic drain of containment #37.02.01 Containment evacuation system

establish and maintain containment #36.02.01 vacuum

provide RCS leakage detection #36.02.03 CNTS level sensors

provides containment level input for #37.01.06 CFDS automatic fill and drain of containment RCS flow sensors

provides RCS flow indication during #70 hot functional testing and power #88 ascension testing PZR level sensors

provides input for PZR level control #33.02.01 Island mode operation

NPMs can operate independently #99 and #100 from offsite transmission grid.

NuScale US460 SDAA 14.2-185 Revision 0

NuScale Final Safety Analysis Report Inspections, Tests, Analyses, and Acceptance Criteria 14.3 Inspections, Tests, Analyses, and Acceptance Criteria 14.3.1 Introduction This section provides guidance regarding the development of Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC). 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 plant license. The successful completion of ITAAC constitutes the basis for the NRC determination to allow operation of a facility certified under 10 CFR 52, and the ITAAC material expires at initial fuel loading.

The ITAAC are used to verify selected as-built top-level design features. A design feature is a physical attribute or performance characteristic of structures, systems, and components (SSC). Not all top-level design features are 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 top-level design features were identified and selected to be verified by ITAAC. The contents of the design commitments 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 ITAAC.

COL Item 14.3-1: An applicant that references the NuScale Power Plant US460 standard design will provide the site-specific selection methodology and inspections, tests, analyses, and acceptance criteria for emergency planning.

COL Item 14.3-2: An applicant that references the NuScale Power Plant US460 standard design 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 Top-Level Design Features and Inspections, Tests, Analyses, and Acceptance Criteria First Principles General criteria that provide clarity on the scope of top-level design features and ITAAC are discussed below. These criteria are consolidated and grouped into two sets: (1) top-level design features scope first principles and (2) ITAAC scope first principles.

A first principles approach is considered such that the ITAAC 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 and the first principles for determining whether a design feature needs an ITAAC. Consistent with these first principles, the selection of the top-level design features is based on the safety significance of SSC, their importance in various safety analyses, and their functions for defense-in-depth considerations.

NuScale US460 SDAA 14.3-1 Revision 0

NuScale Final Safety Analysis Report Inspections, Tests, Analyses, and Acceptance Criteria The first principles for determining the scope of top-level design features and ITAAC are described in Section 14.3.2.1 and Section 14.3.2.2.

14.3.2.1 Top-Level Design Features Scope First Principles Top-level design features are limited to the following:

design features of safety-related SSC

design features of safety-related or nonsafety-related SSC that protect safety-related components

design features of security system physical SSC

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. Furthermore, top-level design features are limited to include the following:

Not all safety-related design features are top-level design features. Refer to Section 14.3.2.1.3 for further discussion.

Not all design features contained in the accident analyses are top-level design features. Refer to Section 14.3.2.1.4 for further discussion.

Operational programs and post-fuel load testing are not design features. Refer to Section 14.3.2.1.5 for further discussion.

Some risk-significant design features identified by the PRA do not need to be specifically addressed because they are indirectly addressed by design features that are addressed by other design commitments. Refer to Section 14.3.2.1.6 for further discussion.

To the extent that SSC are already the subject of a design commitment by reason of a design basis accident mitigation function design feature, 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. Refer to Section 14.3.2.1.7 for further discussion.

Only fixed design features that are installed before fuel loading and that are expected to be in place for the lifetime of the plant are considered for inclusion as top-level design features. Refer to Section 14.3.2.1.8 for further discussion.

Design features of systems with no safety significance are not considered for inclusion.

14.3.2.1.1 Characteristics of Top-Level Design Features The following describes the top-level design features for the NuScale Power Plant US460 standard design.

NuScale US460 SDAA 14.3-2 Revision 0

NuScale Final Safety Analysis Report Inspections, Tests, Analyses, and Acceptance Criteria A design feature is either a physical attribute or a performance characteristic of SSC. The top-level design features contained in the ITAAC are associated with the

reactor coolant pressure boundary.

containment pressure boundary.

Portions of the Seismic Category I Reactor Building and Control Building.

Radioactive Waste Category RW-IIa Radioactive Waste Building.

control room envelope.

safety-related equipment qualification.

nonsafety-related equipment qualification of SSC located within the boundaries of the NuScale Power Module (NPM) that have augmented Seismic Category I or environmental qualification requirements.

nonsafety-related equipment qualification of SSC that have augmented Seismic Category I or environmental qualification requirements and provide one of the following functions:

provides physical support of irradiated fuel provides a path for makeup water to the ultimate heat sink provides containment of the ultimate heat sink water provides monitoring of ultimate heat sink water level

nonsafety-related equipment qualification of SSC classified as RW-IIa used for processing gaseous radioactive waste.

safety-related component performance.

SSC providing protection of safety-related components.

nonsafety-related SSC that perform a credited function in Chapter 15 analyses.

safety-related protection systems (reactor trip and engineered safety features actuation systems).

components providing radiation protection for personnel and safety-related equipment.

new and spent fuel storage.

security system physical components.

Examples of physical attributes included in top-level design features are safety-related equipment qualification, location of fire barriers, and attenuation capabilities of radiation shields.

Examples of performance characteristics included in top-level design features are building seismic performance, safety-related piping conformance to American Society of Mechanical Engineers Code Section III requirements, NuScale US460 SDAA 14.3-3 Revision 0

NuScale Final Safety Analysis Report Inspections, Tests, Analyses, and Acceptance Criteria valve stroke time, and safety-related components' automatic response to the module protection system.

14.3.2.1.2 Severe Accident Design Features The ITAAC are used to verify that the severe accident top-level design features 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 would discuss that the severe accident containment flooding system exists but would not specify the capabilities of associated pumps.

14.3.2.1.3 Safety-Related Design Features Not all safety-related design features are considered top-level design features and do not need to be explicitly addressed. Examples of safety-related component design features that generally do not warrant inclusion are listed below:

instrument lines

fill lines

drains

American Society of Mechanical Engineers 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 Verified by Inspections, Tests, Analyses, and Acceptance Criteria Only the top-level design features are verified by ITAAC. Each ITAAC section includes a table that correlates the top-level design features contained in design commitments with the 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

physical security NuScale US460 SDAA 14.3-4 Revision 0

NuScale Final Safety Analysis Report Inspections, Tests, Analyses, and Acceptance Criteria 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 the FSAR are preserved in ITAAC.

14.3.2.1.5 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 design features and are therefore not included in the ITAAC.

Those aspects of the design that cannot be verified until after fuel loading are not included in ITAAC because 10 CFR 52 requires the ITAAC to be satisfied before fuel loading. For these, the Initial Test Program verifies aspects of the design after fuel load, but before operation. Examples include startup and power ascension test program verifications 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 is a condition of the license.

14.3.2.1.6 Risk-Significant Design Features Some risk-significant design features identified by the PRA do not need to be specifically addressed 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. If equipment is designed and qualified for the seismic design basis, the design process is such that the added capability assumed in the PRA is inherently present.

The risk-significant design features that are included in the design commitments and have associated ITAAC are identified.

14.3.2.1.7 Design Features Necessary or Important to Severe Accident Mitigation There are some SSC that mitigate design basis accidents and provide an important success path for severe accident mitigation. The severe accident mitigation design features that are included in the design commitments and have associated ITAAC are identified.

14.3.2.1.8 Fixed Design Features Installed Before Fuel Loading Those aspects of the design that pertain to portable items or consumables rather than fixed design features are not included. Because hardware such as fuel cannot be installed in the reactor until after completion of the ITAAC and because the fuel is periodically replaced, fuel is not an appropriate topic for ITAAC.

NuScale US460 SDAA 14.3-5 Revision 0

NuScale Final Safety Analysis Report Inspections, Tests, Analyses, and Acceptance Criteria 14.3.2.2 Inspections, Tests, Analyses, and Acceptance Criteria Scope First Principles The following criteria are considered when determining what information warrants inclusion in the ITAAC entries:

The design commitment is extracted directly from the ITAAC design descriptions and differences in text are minimized, unless intentional.

The NRC safety determination is based solely on the FSAR 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 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, can verify one or more provisions in the design commitment, as defined by the ITAAC.

14.3.2.2.1 Design Commitments Only Include Components Required to Perform System Functions in the Inspections, Tests, Analyses, and Acceptance Criteria System Description Not every element of a design commitment specified in ITAAC has a corresponding verification requirement. For example, the safety classification of SSC 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.

14.3.3 Inspections, Tests, Analyses, and Acceptance Criteria Information 14.3.3.1 Inspections, Tests, Analyses, and Acceptance Criteria Tables A table of ITAAC entries is provided for each system that has design commitments in the ITAAC design description. A four-column format for the ITAAC table is used with the first column containing numbers to identify the entries' positions in the table. The remaining three columns of the ITAAC table must be read and interpreted together.

The second 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 ITAAC design description.

NuScale US460 SDAA 14.3-6 Revision 0

NuScale Final Safety Analysis Report Inspections, Tests, Analyses, and Acceptance Criteria The third column of the ITAAC table identifies the proposed method by which the licensee verifies the design commitment described in column 2. 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 can 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 fourth column of the ITAAC table identifies the specific acceptance criteria for the inspections, tests, or analyses described in column 3 that, if met, demonstrate that the licensee has met the design commitments in column 2. 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 top-level design features and preparing a design commitment and corresponding ITAAC table entry for each design feature that satisfies the above selection criteria.

The ITAAC table was completed by selecting the method to be used for verification (i.e., an inspection, a test, an analysis, or a combination of these) and the acceptance criteria the as-built design features are measured against.

Where ITAAC are verified by a preoperational test, the test is established in accordance with the Initial Test Program described in Section 14.2 and Regulatory Guide 1.68. Conversion or extrapolation of test results from the test conditions to design conditions 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. For many of the ITAAC, the NuScale US460 SDAA 14.3-7 Revision 0

NuScale Final Safety Analysis Report Inspections, Tests, Analyses, and Acceptance Criteria acceptance criterion is a statement that the as-built facility has the design feature identified in the design commitment.

A guiding principle for acceptance criteria preparation is the recognition that the criteria are objective and unambiguous. 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 can be more general because the detailed supporting information in the FSAR does not lend itself to concise verification.

In some cases, the ITAAC contain numerical values from the FSAR that are not specifically identified in the design commitment column of the ITAAC table. 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.

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.

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 could indicate 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.

Where appropriate, the detailed design information provided in the FSAR 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.

For each ITAAC table, an accompanying table is provided that includes additional information for each entry in the ITAAC table. This information may include:

locations of detailed design information in the FSAR.

details and references associated with the design commitment.

additional details associated with the conduct of the inspections, tests, and analyses.

details that further define the acceptance criteria.

NuScale US460 SDAA 14.3-8 Revision 0

NuScale Final Safety Analysis Report Inspections, Tests, Analyses, and Acceptance Criteria 14.3.4 Treatment of Module-Specific and Shared Structures, Systems, and Components in Inspections, Tests, Analyses, and Acceptance Criteria 14.3.4.1 Inspections, Tests, Analyses, and Acceptance Criteria for Module-Specific Structures, Systems, and Components Module-specific SSC are specific to and support operation of a single NPM. The module-specific design is identical between NPMs. If a design feature of a module-specific SSC meets the first principles for identification as a top-level design feature and verification through ITAAC as described in Section 14.3.2, then its ITAAC are entered into its associated system's ITAAC table. Chapter 2 of ITAAC includes an entry for each module-specific system that is either fully or partially within the scope of the NuScale Power Plant US460 standard design.

The ITAAC of a given module-specific system are the same for all NPMs and are only recorded once in an ITAAC table.

However, each ITAAC for a given module-specific system must be completed for each NPM.

14.3.4.2 Inspections, Tests, Analyses, and Acceptance Criteria for Shared Structures, Systems, and Components Shared SSC support multiple NPMs. If a design feature of a shared SSC meets the first principles for identification as a top-level design feature and verification through ITAAC as described in Section 14.3.2, then its ITAAC are entered into its associated system's ITAAC table. Chapter 3 of ITAAC includes an entry for each shared system that is either fully or partially within the scope of the NuScale Power Plant standard design. Additionally, shared ITAAC address non-SSC design activities that are applicable to more than one system or NPM such as human factors engineering.

Shared systems that must be completed to support the operation of the first NPM have their ITAAC completed once. If shared systems require a portion of the system to be completed to support the operation of the first NPM, then the applicable ITAAC is in a module-specific ITAAC table and must be completed for each associated NPM.

NuScale US460 SDAA 14.3-9 Revision 0

ML22362A070

Text

SUBJECT:

REFERENCES:

Navigation menu

Search