ML20197A387

Part 02 - Final Safety Analysis Report (Rev. 4.1) - Part 02 - Tier 02 - Chapter 01 - Introduction and General Description of the Plant - Sections 01.01 - 01.10 (Rev. 4.1)

ML20197A387
Person / Time
Site:	NuScale
Issue date:	06/19/2020
From:	Bergman T NuScale
To:	Office of Nuclear Reactor Regulation
Cranston G
References
NUSCALESMRDC, NUSCALESMRDC.SUBMISSION.12, NUSCALEPART02.NP, NUSCALEPART02.NP.5
Download: ML20197A387 (350)

Text

NuScale Standard Plant Design Certification Application Chapter One Introduction and General Description of the Plant PART 2 - TIER 2 Revision 4.1 June 2020

©2020, NuScale Power LLC. All Rights Reserved

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

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

APTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF THE PLANT . . . . . . . . . . .1.1-1 1.1 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-1 1.1.1 Plant Location . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-1 1.1.2 Containment Type . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-2 1.1.3 Reactor Type . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-2 1.1.4 Power Output . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-2 1.1.5 Schedule . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-2 1.1.6 Format and Content . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-2 1.2 General Plant Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1 1.2.1 Principal Site Characteristics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-1 1.2.2 General Arrangement of Major Structures and Equipment . . . . . . . . . . . . . . . . . . 1.2-11 1.2.3 Plant Features of Special Interest . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-18 1.3 Comparison with Other Facilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-1 1.4 Identification of Agents and Contractors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.4-1 1.4.1 Not Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.4-1 1.4.2 Division of Responsibility . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.4-1 1.4.3 Principal Consultants and Other Participants. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.4-1 1.5 Requirements for Additional Technical Information. . . . . . . . . . . . . . . . . . . . . . . . . . . 1.5-1 1.5.1 NuScale Testing Programs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.5-1 1.5.2 NuScale Test and Inspection Plans. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.5-7 1.6 Material Referenced . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.6-1 1.7 Drawings and Other Detailed Information. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-1 1.7.1 Electrical and Instrumentation and Control Drawings . . . . . . . . . . . . . . . . . . . . . . . . 1.7-1 1.7.2 Piping and Instrumentation Diagrams . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-1 1.8 Interfaces with Certified Design . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.8-1 1.8.1 Combined License Information Items. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.8-1 1.8.2 Departures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.8-1 1.9 Conformance with Regulatory Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-1 1.9.1 Conformance with Regulatory Guides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-1 1.9.2 Conformance with Standard Review Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-2 1.9.3 Generic Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-2 1.9.4 Operational Experience (Generic Communications) . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-2 2 i Revision 4.1

1.9.5 Advanced and Evolutionary Light-Water Reactor Design Issues. . . . . . . . . . . . . . . 1.9-3 1.9.6 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-3 1.10 Nuclear Power Plants to be Operated on Multi-Unit Sites . . . . . . . . . . . . . . . . . . . . . 1.10-1 2 ii Revision 4.1

le 1.1-1: Acronyms and Abbreviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.1-4 le 1.2-1: Overall Characteristics of a NuScale Power Plant . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-19 le 1.2-2: Design Features of a NuScale Power Module . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-20 le 1.3-1: NuScale Plant Comparison with Other Facilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-2 le 1.3-2: Safety Systems and Components Required to Protect the Reactor Core - NuScale Comparison with Other Facilities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.3-3 le 1.6-1: NuScale Referenced Topical Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.6-2 le 1.6-2: NuScale Referenced Technical Reports. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.6-3 le 1.7-1: Instrumentation and Controls Functional and Electrical One-Line Diagrams . . . . . 1.7-2 le 1.7-2: System Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-3 le 1.8-1: Summary of NuScale Certified Design Interfaces with Remainder of Plant. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.8-2 le 1.8-2: Combined License Information Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.8-3 le 1.9-1: Conformance Status Legend . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-4 le 1.9-2: Conformance with Regulatory Guides . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-5 le 1.9-3: Conformance with NUREG-0800, Standard Review Plan (SRP) and Design Specific Review Standard (DSRS). . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-46 le 1.9-4: Conformance with Interim Staff Guidance (ISG) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-196 le 1.9-5: Conformance with TMI Requirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-206 le 1.9-6: Evaluation of Operating Experience (Generic Letters and Bulletins) . . . . . . . . . . . 1.9-216 le 1.9-7: Conformance with Advanced and Evolutionary Light Water Reactor Design Issues (SECYs and Associated SRMs) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.9-218 le 1.9-8: Conformance with SECY-93-087, "Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor Designs" . . . . . . 1.9-220 2 iii Revision 4.1

re 1.2-1: Conceptual Site Layout. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-21 re 1.2-2: NuScale Functional Boundaries . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-22 re 1.2-3: Schematic of a Single NuScale Power Module and Associated Secondary Equipment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-23 re 1.2-4: Layout of a Multi-Module NuScale Power Plant. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-24 re 1.2-5: Cutaway Illustration of 12 Module Configuration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-25 re 1.2-6: Cutaway View of NuScale Power Module. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-26 re 1.2-7: Steam Generator and Reactor Flow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-27 re 1.2-8: Decay Heat Removal System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-28 re 1.2-9: Emergency Core Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-29 re 1.2-10: Reactor Building 24-0 Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-30 re 1.2-11: Reactor Building 35'-8" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-31 re 1.2-12: Reactor Building 50'-0" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-32 re 1.2-13: Reactor Building 62'-0" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-33 re 1.2-14: Reactor Building 75'-0" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-34 re 1.2-15: Reactor Building 86'-0" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-35 re 1.2-16: Reactor Building 100'-0" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-36 re 1.2-17: Reactor Building 126'-0" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-37 re 1.2-18: Reactor Building 145'-6" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-38 re 1.2-19: Reactor Building East and West Section View. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-39 re 1.2-20: Reactor Building South Section View. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-40 re 1.2-21: Control Building 50'-0" Elevation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-41 re 1.2-22: Control Building 63'-3" Elevation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-42 re 1.2-23: Control Building 76'-6" Elevation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-43 re 1.2-24: Control Building 100'-0" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-44 re 1.2-25: Control Building 120'-0" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-45 re 1.2-26: Control Building North Section View . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-46 re 1.2-27: Control Building West Section View. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-47 re 1.2-28: Radioactive Waste Building 71'-0" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-48 re 1.2-29: Radioactive Waste Building 82'-0" Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-49 re 1.2-30: Radioactive Waste Building 100'-0" Elevation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-50 re 1.2-31: Radioactive Waste Building 120-0 Elevation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-51 2 iv Revision 4.1

re 1.2-32: Radioactive Waste Building North and South Section Views . . . . . . . . . . . . . . . . . . . 1.2-52 re 1.2-33: Radioactive Waste Building West Section View . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.2-53 re 1.7-1a: Electrical Symbols . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-5 re 1.7-1b: Electrical Symbols . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-6 re 1.7-2: Instrumentation and Controls Symbol Legend . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-7 re 1.7-3a: Piping and Instrumentation Diagram Legends . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-8 re 1.7-3b: Piping and Instrumentation Diagram Legends . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-9 re 1.7-3c: Piping and Instrumentation Diagram Legends . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-10 re 1.7-3d: Piping and Instrumentation Diagram Legends . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-11 re 1.7-3e: Piping and Instrumentation Diagram Legends . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-12 re 1.7-3f: Piping and Instrumentation Diagram Legends . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.7-13 2 v Revision 4.1

Introduction This document represents the Final Safety Analysis Report (FSAR) required under 10 CFR 52.47(a) to be provided as part of an application for a standard design certification under 10 CFR 52, Subpart B and will be referred to as such throughout. It describes the NuScale Power, LLC design, including (1) the design bases and limits on its operation; (2) a safety analysis of the structures, systems, and components and of the facility as a whole; and (3) the information prescribed in 10 CFR 52.47(a) that is relevant to the NuScale design.

A NuScale Power Module (NPM) shown in Figure 1.2-6 and Figure 1.2-7, is a collection of systems, sub-systems, and components that together constitute a modularized, movable, nuclear steam supply system (NSSS). The NPM is composed of a reactor core, a pressurizer, and two steam generators integrated within a reactor pressure vessel (RPV) and housed in a compact steel containment vessel.

The NuScale advanced small modular reactor plant design is scalable, such that from one (1) to twelve (12) NPMs operate within a single Reactor Building. The information provided in this FSAR includes the design of an individual NPM, as well as plant design and interfaces for a 12 NPM facility. In general, chapters describe a single module. Multi-module information is only noted where warranted (e.g., shared systems or analyses such as seismic).

The NuScale design features:

• No AC or DC power required for safe shutdown and cooling

• Compact helical coil steam generators with reactor pressure on the outside of the tubes

• High-strength steel containment immersed in a pool of water

• Sub-atmospheric containment pressure during normal operation

• Small core with a correspondingly small source term

• Comprehensive digital instrumentation and controls (I&C) monitoring and control Important features of a multi-unit plant include:

• a scalable plant design, which allows for incremental plant capacity growth.

• a compact nuclear island.

• the ability to operate in "island mode".

1 Plant Location The NuScale Power Plant is designed to be located on a site having site characteristics (e.g.,

seismology, hydrology, meteorology, geology, and other site-related characteristics) bounded by the parameters described in Chapter 2, Site Characteristics.

Item 1.1-1: A COL applicant that references the NuScale Power Plant design certification will identify the site-specific plant location.

2 1.1-1 Revision 4.1

The NuScale containment vessel (CNV) is a supported, cylindrical vessel-type containment that is designed to withstand limiting high-pressure transients. The containment vessel (CNV) is an American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Class MC (steel) containment that is designed, analyzed, fabricated, inspected, tested and stamped as an ASME BPVC Class 1 pressure vessel. The CNV internal pressure is maintained at a vacuum during normal operations and as such insulation materials are not required between the reactor vessel and the CNV. The containment vessels are mounted to the Reactor Building module compartment walls and at the bottom within the Reactor Building pool.

3 Reactor Type The NuScale NSSS is a passive NuScale-designed small modular pressurized water reactor.

This design is comprised of an integral power module consisting of a reactor core, two steam generator tube bundles, and a pressurizer contained within a single reactor vessel, along with the containment vessel that immediately surrounds the reactor vessel. This design eliminates the need for external piping to connect the steam generators and pressurizer to the RPV. Natural circulation provides reactor coolant system flow, thereby eliminating the need for reactor coolant pumps.

4 Power Output A NuScale Power Plant consists of from one to12 NPMs. Each NPM is rated at 160 MWt (1,920 MWt, total), with approximately 50 MWe (600 MWe total) output. Electrical output is dependent on environmental conditions. When considering house loads, the total net output is approximately 570 MWe for a 12 NPM facility. Design power assumes an additional 2 percent to account for measurement uncertainty.

5 Schedule Item 1.1-2: A COL applicant that references the NuScale Power Plant design certification will provide the schedules for completion of construction and commercial operation of each power module.

6 Format and Content 6.1 Regulatory Guide 1.206 The format and content of this FSAR generally follow the format and content guidelines of Regulatory Guide 1.206. However, where applicable, Sections may be skipped or additional sections inserted. In addition, this FSAR includes Chapter 20, Mitigation of Beyond-Design-Basis Events and Chapter 21, Multi-Module Design Considerations, which are not included in Regulatory Guide (RG) 1.206.

6.2 Standard Review Plan - NuScale Design Specific Review Standard A NuScale design specific review standard (DSRS) has been developed by the NRC as a supplement to NUREG-0800, Standard Review Plan for the Review of Safety Analysis 2 1.1-2 Revision 4.1

NuScale design. A detailed evaluation of conformance with the NuScale DSRS and the SRP is provided in Section 1.9.

6.3 Text, Tables, and Figures Tables and figures are typically identified by the "X.Y" section in which they appear and are numbered sequentially. For example, Table 1.1-1 and Figure 1.1-1 would be the first table and figure appearing in Section 1.1. Figures consist of diagrams, plots, pictures, graphs, or other illustrations. Tables and figures are located at the end of the applicable "X.Y" section immediately following the text. The exception to this is for large "X.Y.Z" sections, in which the tables and figures are numbered sequentially in that section. For example, Table 3.9.3-1 and Figure 3.9.3-1 would be the first table and figure appearing in Section 3.9.3. Again, the tables and figures are located at the end of the applicable section intermediately following the text.

6.4 Page Numbering Section pages are numbered sequentially and are typically identified by the "X.Y" section followed by a sequential number. The exception to this convention is for chapter appendices, which are numbered by the chapter number and appendix letter followed by a sequential number. For example, 3A-1 is the first page of Appendix A to Chapter 3.

6.5 Proprietary Information This FSAR does not contain proprietary or safeguards information. Some portions of this FSAR are classified as sensitive and withheld from public disclosure pursuant to 10 CFR 2.390 and Regulatory Issue Summary (RIS) 2005-26. Such material is clearly marked and provided with the non-public version of the FSAR. A separate public version of the FSAR is provided that removes the withheld material. Proprietary or safeguards information that is necessary for the complete review of the design certification is provided to the NRC separately in the form of topical or technical reports. Topical and technical reports that are incorporated by reference are listed in Tables 1.6-1 and 1.6-2, respectively.

6.6 Acronyms and Abbreviations A list of acronyms and abbreviations used in this FSAR is provided in Table 1.1-1, Acronyms and Abbreviations.

2 1.1-3 Revision 4.1

nym or Description reviation alternate AC power S auxiliary AC power source auxiliary boiler system S Annex Building HVAC system R Advanced Boiling Water Reactor alternating current American Concrete Institute Availability Controls Manual S Advisory Committee on Reactor Safeguards Atomic Energy Act air filtration unit S auxiliary feedwater system authority having jurisdiction air handling unit Authorized Inspection Agency American Institute of Steel Construction American Iron and Steel Institute RA as low as reasonably achievable actuation logic unit R advanced light water reactor A Air Movement and Control Association International, Inc.

Annex Building American Nuclear Society American National Standards Institute axial offset axial offset anomaly anticipated operational occurrence air-operated valve American Petroleum Institute R Advanced Pressurized Water Reactor augmented quality area radiation monitor all rods out acceleration response spectra E American Society of Civil Engineers adjustable speed drive RAE American Society of Heating, Refrigerating, and Air-Conditioning Engineers American Society for Metals International E American Society of Mechanical Engineers M American Society for Testing and Materials Administration and Training Building S anticipated transient without scram all-volatile treatment American Welding Society A American Water Works Association boron addition system T boric acid storage tank E beyond design basis event EE beyond design basis external event 2 1.1-4 Revision 4.1

nym or Description reviation backup diesel generator beginning of cycle beginning of life balance-of-plant S balance-of-plant drain system bioprocessing equipment backup power supply system C Boiler and Pressure Vessel Code Ballistic Research Laboratory battery room ventilation system Branch Technical Position boiling water reactor continuous air monitor S condenser air removal system central alarm station compressed air system E common cause basic event F conditional core damage frequency P conditional core damage probability common cause failure counter current flow limitation P conditional containment failure probability core damage event core damage frequency conceptual design information certified design material T core damage source term control element assembly containment evacuation system containment event tree S central and eastern United States computational fluid dynamics S containment flooding and drain system Code of Federal Regulations containment flange tool critical heat flux R critical heat flux ratio S condensate and feedwater system S containment heat removal system S chilled water system T containment integrated leak rate test civil interface macro clean-in-place containment isolation system containment isolation valve conditional large release frequency containment leakage rate testing A Crane Manufacturers Association of America code management software R certified material test report 2 1.1-5 Revision 4.1

nym or Description reviation S containment system containment vessel F containment vessel failure certificate of compliance combined license A combined license application R core operating limits report S communication system S condensate polisher resin regeneration system condensate polishing system complete quadratic combination control rod assembly Control Building M control rod drive mechanism S control rod drive system control room envelope S control room habitability system control rod misoperation S normal control room HVAC system core support assembly RS certified seismic design response spectra RS-HF certified seismic design response spectra - high frequency containment sampling system condensate storage tank combustion turbine generator Central Utility Building P Comprehensive Vibration Assessment Program S chemical and volume control system circulating water system diversity and defense in depth design acceptance criteria distributed antenna system diverse actuation system dry active waste design basis accident design basis event B design basis pipe break T design basis source term design basis tornado direct current Design Certification Application Design Control Document (Note - this is synonymous with FSAR in this document) direct containment heating distributed control system distributed Doppler coefficient dry dock gate Diesel Generator Building VS Diesel Generator Building HVAC system S decay heat removal system display interface module 2 1.1-6 Revision 4.1

nym or Description reviation dimethylamine departure from nucleate boiling R departure from nucleate boiling ratio Department of Energy Department of Transportation P Design Reliability Assurance Program S Design Specific Review Standard digital safety system dry solid waste Doppler temperature coefficient, fuel temperature coefficient, Doppler coefficient demineralized water system exclusion area boundary Emergency Action Level S emergency core cooling system effluent concentration limit equivalent dead load G extensive damage mitigation guidelines S normal DC power system S highly reliable DC power system S-C EDSS-common S-MS EDSS-module-specific engineering design verification S equipment and floor drainage system effective full-power days effective full-power years S 13.8 kV and switchyard system equipment interface module low voltage AC electrical distribution system R evolutionary light water reactor electromagnetic compatibility AP evaluation model development and assessment process M electromagnetic drive mechanism electromagnetic interference S medium voltage AC electrical distribution system end of cycle emergency operations facility end of life emergency operating procedure electrical penetration assembly Environmental Protection Agency emergency procedure guidelines Electric Power Research Institute emergency planning zone equipment qualification P equipment qualification data package F equipment qualification record file A Energy Research and Development Administration S emergency response data system emergency response facility Emergency Response Organization 2 1.1-7 Revision 4.1

nym or Description reviation equipment requirement specification emergency safeguards actuation system WR Economic Simplified Boiling Water Reactor engineered safety feature S engineered safety features actuation system equivalent static load early site permit ethanolamine Electrical Transient Analyzer Program functional analysis flow-accelerated corrosion factory acceptance test fracture appearance transition temperatures fuel-coolant interaction fan coil unit final design approval fire detection system Federation Europeenne de la Manutention Federal Energy Regulatory Commission fitness-for-duty fast Fourier transform fire hazards analysis fuel handling equipment fuel handling machine foundation input response spectra flow-indicating transmitter flow-induced vibration diverse and flexible coping strategies (based on NRCs Fukushima task force recommendations)

A flooding probabilistic risk assessment A failure modes and effects analysis K first-of-a-kind figure of merit A field programmable gate array Fire Protection Program fire probabilistic risk assessment fire protection system functional requirements analysis fiber-reinforced polymer Final Safety Analysis Report (Note - this is synonymous with DCD in this document)

FLEX support guidelines fluid-structure interaction fire safe shutdown analysis fire safe shutdown Fussell-Vesely feedwater Fire Water Building feedwater heater feedwater isolation valve B feedwater line break B feedwater pipe break 2 1.1-8 Revision 4.1

nym or Description reviation V feedwater regulating valve feedwater system S feedwater treatment system granulated activated charcoal General Design Criteria S grounding and lightning protection system S ground motion response spectra graded quality assurance S gaseous radioactive waste system generic safety issue W gas tungsten arc weld generic technical specifications heat-affected zone PF high confidence of low probability of failure high-conductivity waste Hardware Development Plan E high-density polyethylene human engineering discrepancy Heat Exchanger Institute high-energy line break human error probability A high-efficiency particulate air human factors engineering, human failure events TS human factors engineering issue tracking system hot full power high integrity container highly integrated protection system E heavy load handling equipment human machine interface hydraulic-operated valve high pressure, horsepower WH high pressure feedwater heater human performance monitoring E high pressure melt ejection health physics network human reliability analysis hardware requirement specification human-system interface C heating ventilation and air conditioning S feedwater heater vents and drains system M hard-wired module hot zero power instrumentation and controls inadvertent actuation block instrument air system International Building Code in-core instrumentation system integrated control system inside diameter interface design description 2 1.1-9 Revision 4.1

nym or Description reviation infrequent event, initiating event International Electrotechnical Commission Institute of Electrical and Electronics Engineers Illuminating Engineering Society of North America integral effects test C intergranular stress-corrosion cracking important human action integrated leak rate testing Idaho National Laboratory Institute of Nuclear Power Operations S inadvertent opening of the turbine bypass system implementation plan intermediate pressure WH intermediate pressure feedwater heater integrated safety analysis, Instrument Society of America interim staff guidance inservice inspection inservice leak and hydro CA interfacing systems loss-of-coolant accident independent support motion International Organization for Standardization in-structure response spectra inservice testing integrated system validation C Inspections, Tests, Analyses, and Acceptance Criteria inspection, testing, and maintenance Initial Test Program in-vessel retention Japan Lessons-Learned Directorate leak-before-break limiting condition for operation local control station low-conductivity waste Licensee Event Report R linear heat generation rate local leak rate test A loss-of-coolant accident A loss of large areas P loss of offsite power low pressure WH low pressure feedwater heater low power and shutdown low population zone lower riser assembly large release frequency liquid ring vacuum pump liquid radioactive waste S liquid radioactive waste system, liquid radwaste system level switch, high level switch, low 2 1.1-10 Revision 4.1

nym or Description reviation limiting safety system setting load manual tap changers long-term core cooling P low temperature overpressure protection S loss of normal access to the ultimate heat sink S liquid waste management system light water reactor module assembly equipment main condenser motor control center FR minimum critical heat flux ratio main control room module control system R module critical year master equipment list S meteorological and environmental monitoring system main feedwater maximum hypothetical accident module heatup system monitoring and indication bus microbiologically induced corrosion Massachusetts Institute of Technology module lifting adapter master logic diagram multiple, multi-module AF multi-module adjustment factor multi-module issue SF multi-module performance shaping factor moment magnitude scale middle of cycle motor-operated valve module protection system main power transformer magnetic speed pickup main steam isolation V main steam isolation bypass valves main steam isolation valve B main steam line break multiple spurious operations B main steam pipe break I mitigating system performance index main steam system V main steam safety valve moderator temperature coefficient metric tons, uranium e megawatt electric S maintenance workstation megawatt thermal Not Applicable non-destructive examination 2 1.1-11 Revision 4.1

nym or Description reviation nitrogen distribution system non-destructive testing Nuclear Energy Institute C North American Electric Reliability Corporation new fuel assembly new fuel elevator new fuel jib crane A National Fire Protection Association network interface controller National Institute of Standards and Technology

-1 NuScale Integral System Test Facility neutron monitoring system nuclear overhead and gantry NuScale Power Module NuScale Power Plant nominal pipe size H net positive suction head Nuclear Regulatory Commission nuclear reliability factor neutron source assembly C Nuclear Safety Analysis Center S nuclear steam supply system F Near-Term Task Force operating basis earthquake operational condition sampling outside diameter overspeed detection circuit M Offsite Dose Calculation Manual operating experience operating experience review HS overhead heavy load handling system L Oak Ridge National Laboratory P Operational Radiation Protection Program operational support center A Occupational Safety and Health Administration overspeed protection system Oregon State University piping and instrumentation diagram protected area A public address/general alarm S priority actuation and control system post-accident monitoring private branch exchange primary coolant activity V prestressed concrete containment vessel Process Control Program plant control system peak cladding temperature S pool cleanup system power distribution center 2 1.1-12 Revision 4.1

nym or Description reviation principal design criteria power dependent insertion limit differential pressure indicating transmitter process feed tank peak ground acceleration concentration of H+ ion on a logarithmic scale (temperature dependent) proportional integral derivative particulate, iodine, and noble gas phenomena identification and ranking table pressure indicating transmitter programmable logic controller pool leakage detection D programmable logic design description P Programmable Logic Development Plan S pool leakage detection system GR peak linear heat generation rate priority logic module programmable logic requirement specification plant lighting system P Programmable Logic Verification and Validation Plan probable maximum flood probable maximum precipitation V power-operated relief valve plant operating state power-operated valve personnel protective equipment plant protection system probabilistic risk assessment pressure relief valve V primary system containment isolation valves pool surge control system power spectra density S power supply monitoring system process sampling system phase separator tank public switched telephone network C performance and test acceptance criteria band pressurized thermal shock polyvinyl chloride S plant-wide video monitoring system T post-weld heat treatment pressurized water reactor potable water system CC primary water stress-corrosion cracking pressurizer quality assurance Quality Assurance Program D Quality Assurance Program Description quadrant power difference quick disconnect 2 1.1-13 Revision 4.1

nym or Description reviation quadrant power fractions request for additional information Reliability Assurance Program risk achievement worth Reactor Building crane M Reactor Building components Reactor Building HVAC system radiologically controlled area A rod control cluster assembly WS reactor component cooling water system reactor coolant pump B reactor coolant pressure boundary A Resource Conservation and Recovery Act reactor coolant system reactor drain tank rod ejection accident Radiological Effluent Technical Specifications radio frequency interference refueling pool reactor flange tool Regulatory Guide residual heat removal regenerative heat exchanger regulatory issue summary review level earthquake radiation monitoring fixed area radiation monitoring system S risk-managed technical specifications reverse osmosis Reactor Oversight Process S reactor pool cooling system rod position indication reactor protection system reactor pressure vessel required response spectrum reactor recirculation valve remote shutdown area results summary report remote shutdown station reactor safety valve reactor trip breaker resistance temperature detector requirements traceability matrix T reference temperature for nil-ductility transition SS regulatory treatment of nonsafety systems rated thermal power S reference temperature, pressurized thermal shock reactor trip system reactor vessel internals reactor vent valve 2 1.1-14 Revision 4.1

nym or Description reviation Radioactive Waste Building CR Radioactive Waste Building control room VS Radioactive Waste Building HVAC system S radioactive waste drain system S radioactive waste management system S raw water supply system Reactor Building reactor core staffing and qualifications DL specified acceptable fuel design limit seismic anchor motion DA severe accident mitigation design alternative G severe accident management guideline Safety Analysis Report secondary alarm station service air system site acceptance testing C smooth bounding analysis curve subscale boundary layer boiling CA small-break loss-of-coolant accident scheduling and bypass module station blackout Security Building HVAC system Seismic Category I Seismic Category II I Seismic Category III Security Buildings stress corrosion-cracking F seismic core damage frequency silicon controlled rectifier SCRAM load secondary sampling system S site cooling water system safety data bus safety display and indication system shutdown margin E secure development and operational environment F single-degree-of-freedom software development process site drainage system Security Building plant security system Secretary of the Commission, Office of the NRC Structural Engineering Institute seismic equipment list Safety Evaluation Report spent fuel assembly safety function module spent fuel pool S spent fuel pool cooling system 2 1.1-15 Revision 4.1

nym or Description reviation spent fuel storage system separation group steam generator strain gauge safeguards information steam generator system steam generator tube failure safety information and control system software integrity level steam line break site layout plan single module seismic margin assessment CNA Sheet Metal and Air Conditioning Contractors' National Association subject matter expert small modular reactor seismic monitoring system Sandia National Laboratories special nuclear material A security owner controlled area solenoid-operated valve R standardized plant analysis risk D self-powered neutron detector security power system P seismic qualification data package F seismic qualification record form G Seismic Qualification Utility Group surveillance requirement standard radiological effluent control Stanford Research Institute staff requirements memorandum Standard Review Plan square root of the sum of the squares spent resin storage tank sump recirculation valve S solid radioactive waste system safe shutdown analysis structures, systems, and components V secondary system containment isolation valve safe shutdown earthquake soil-structure interaction or secondary system isolation secondary sampling system structure-soil-structure interaction station service transformer System-Theoretic Process Analysis SI sensitive unclassified non-safeguards information schedule and voting module service water intake structure S solid waste management system shear wave velocity 2 1.1-16 Revision 4.1

nym or Description reviation D switchyard system task analysis top of active fuel Turbine Building crane turbine bypass system Turbine Building HVAC system thermocouple temperature control unit total dynamic head total dissolved solids E total effective dose equivalent Turbine Generator Building turbine generator system turbine gland sealing system total harmonic distortion treatment of important human actions temperature indicating transmitter thermoluminescent dosimeter SS turbine lube oil storage system Three Mile Island triple module redundancy nil ductility temperature top of concrete test response spectrum technical specifications technical support center Technical Specification Task Force tubular ultrafiltration unit auxiliary transformer W uncontrolled control rod assembly withdrawal at power WS uncontrolled control rod assembly withdrawal from a subcritical or low power or startup condition uniform Doppler coefficient ultimate heat sink uninterruptible power supply Utility Requirements Document uniform response spectrum upper riser assembly/section S United States Geological Survey unresolved safety issue uniform support motion coordinated universal time utility water system verification and validation video display unit vibration indicating transmitter vented lead-acid valve-regulated lead-acid voltage regulating transformer watchdog timer CR waste management control room 2 1.1-17 Revision 4.1

nym or Description reviation wet solid waste Waste Treatment Building Zone of Influence zero period acceleration 2 1.1-18 Revision 4.1

This section summarizes the plant design and provides a general description of the overall facility. The description includes:

• principal design criteria, operating characteristics, and safety considerations

• engineered safety features (ESFs) and emergency systems

• instrumentation, controls, and electrical systems

• power conversion system

• fuel, fuel handling, and storage systems

• plant cooling water systems

• radioactive waste management systems

• auxiliary systems (e.g., compressed air, non-radioactive drains, water systems)

Each COL Applicant will develop a Final Safety Analysis Report (FSAR) that incorporates by reference the NuScale FSAR. The NuScale FSAR includes COL items that identify where site-specific information must be provided. However, in some instances, representative information is necessary to provide context for interface requirements as specified in 10 CFR 52.47(a)(24) and 10 CFR 52.47(a)(25). This representative or conceptual design information (CDI) is outside the scope of the NuScale Power Plant certified design. Where provided, CDI is delineated by double brackets ((())). The scope of the certified design and site-specific design is shown in Figure 1.2-2. The basic systems associated with power generation are shown in Figure 1.2-3. Although some components from these systems are physically located in buildings that are CDI, the system itself is not, with the exception of the clouded portion, which identifies the CDI cooling towers and certain circulating water systems.

Security-related information is delineated using double braces ((. This information is withheld in accordance with 10 CFR 2.390(d)(1). 1 Principal Site Characteristics Figure 1.2-1 presents a representative conceptual layout of the overall site. The majority of the site buildings are located within the protected area (PA) and surrounded by a double fence and intrusion-detection equipment. The PA is located within the security owner controlled area (SOCA) surrounded by an additional single fence. An administration and training building and a warehouse are shown outside of the SOCA fence. A NuScale Power Module (NPM) shown in Figure 1.2-6, is a collection of systems, sub-systems, and components that make up a modularized, movable, nuclear steam supply system (NSSS). Each NPM is comprised of a reactor core, a pressurizer, and two steam generators (SGs) integrated within a reactor pressure vessel (RPV) and housed in a compact steel containment vessel (CNV). The NuScale Power Plant is designed for 1 to 12 NPMs with the associated primary and secondary systems and components necessary to produce power and maintain the facility. This includes main steam systems, turbine generator sets, condensate and feedwater systems, and shared external cooling water systems (Figure 1.2-3), plus module assembly equipment, fuel handling equipment, turbine maintenance equipment, and radioactive 2 1.2-1 Revision 4.1

The following structures are included in the NuScale certified design (Figure 1.2-1 and Figure 1.2-2):

1) Reactor Building (RXB): located above and below grade, houses the following facilities (among others that are not specifically discussed in this section):

• ultimate heat sink (reactor pool, refuel pool, and spent fuel pool)
• fuel handling areas
• remote shutdown station
• primary systems Additional details of the RXB are provided in Section 1.2.2.1.

2) Control Building (CRB): located above and below grade, adjacent to the RXB, provides space for the following facilities:

• main control room (MCR): located below grade, houses the equipment, controls, and indications for operation of the NPMs
• technical support center-located above the MCR, outside the radiological controlled area, provides space to support emergency operations and personnel Additional details of the CRB are provided in Section 1.2.2.2.

3) Radioactive Waste Building (RWB): located above and below grade, provides space for heating ventilating and air conditioning (HVAC) equipment; and radioactive waste treatment and storage equipment. Additional details of the RWB are provided in Section 1.2.2.3.

The following structures are discussed as CDI (Figure 1.2-1 and Figure 1.2-2):

1) Turbine Generator Buildings (TGBs): house the turbine generators and associated equipment. Additional details of the TGBs are provided in Section 1.2.2.5.1.

2) Annex Building (ANB): controls access into the radiologically controlled area (RCA) and provides space for health physics facilities, servicing potentially radioactive and non-radioactive tooling, fixtures, and instrumentation, security services, and various personnel services. Additional details of the ANB are provided in Section 1.2.2.5.2.

3) Security Buildings (SCBs): provide for controlled access into the SOCA and the PA of the plant. Additional details of the SCBs are provided in Section 1.2.2.5.3.

4) Central Utility Building (CUB): houses various equipment for the chilled water system and other ancillary equipment for balance of plant systems. Additional details of the CUB are provided in Section 1.2.2.5.4.

5) Diesel Generator Buildings (DGBs): house the backup diesel generators and associated equipment. Additional details of the DGBs are provided in Section 1.2.2.5.5.

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equipment design and operation are site specific. Additional details of the SCWS are provided in Section 1.2.1.6. 1.1 Facility Description Process Overview The reactor core is located in a core support assembly, which is seated in the lower RPV assembly. A central hot leg riser is connected to the top of the core support assembly. The reactor core transfers heat into the reactor coolant and the heated reactor coolant flows upward through the core and lower and upper riser assemblies. The heated coolant exits the upper riser assembly and is redirected downwards into the SG region between the vessel wall and the upper riser assembly. As the reactor coolant transfers heat to the SGs, it cools and becomes denser, which drives the natural circulation flow. The coolant returns to the bottom of the vessel through the downcomer and back into the reactor core, where the cycle begins again (Figure 1.2-7). On the secondary side, preheated feedwater is pumped into the tube side of the SGs where it boils. As the steam flows upward in the tubes, it is continually heated to produce superheated steam before exiting the top of the SGs. The superheated steam is directed to a dedicated steam turbine. A generator, driven by the turbine, creates electric power that is delivered to the utility grid through a step-up transformer. A turbine bypass line provides up to 100 percent of the rated main steam flow directly from the associated steam generators to the main condenser in a controlled manner to remove heat from the reactor following a load reduction or loss of electrical load. Steam that exits or bypasses the turbine is directed to the condenser. A shared circulating water loop removes heat and condenses the steam for up to 6 condensers. The condensate is pumped through condensate polishing equipment to the inlet of the variable speed feedwater pumps. A small amount of steam is extracted from turbine stages to preheat the feedwater and increase plant efficiency. Feedwater regulating valves control feed flow into the SGs. ((Heat from the circulating water loop from up to 6 condensers is rejected to atmosphere by a set of evaporative mechanical-draft cooling towers. Two sets of cooling towers are provided for 12 NPMs.)) 1.1.1 Principal Design Criteria The design provides a simple, safe reactor and provides the following:

• reliable, passive safety systems that are simple in design and operation, and are not reliant on electrical power to fulfill their safety functions
• safety features that assure a core damage frequency significantly lower than the current light water reactor fleet 2 1.2-3 Revision 4.1
• modularization to enable in-shop fabrication of reactor and containment components 1.1.2 Operating Characteristics The NPM is designed to operate up to full power conditions using natural circulation as the means of providing reactor coolant flow, eliminating the need for reactor coolant pumps.

The NPMs are partially immersed in a reactor pool and protected by passive safety systems. Each NPM has a dedicated emergency core cooling system (ECCS) and decay heat removal system (DHRS). Important features of the NPM include the following:

• a small, modular design
• an integral pressurized water reactor (PWR) NSSS that combines the reactor core, SGs, and pressurizer within the RPV, eliminating the need for external piping to connect the SGs and pressurizer to the RPV
• natural circulation provides the driving force for reactor coolant flow, eliminating the need for reactor coolant pumps
• an RPV housed in a steel containment partially immersed in water, providing an effective passive heat sink for long-term decay heat removal
• a steel containment operated at a vacuum, eliminating the need for insulation on the RPV
• passive safety systems that are not reliant on electrical power Table 1.2-1 presents the overall characteristics of the NuScale Power Plant.

The NPM is designed to perform normal power maneuvers. Electric power can be adjusted with turbine bypass to the condenser. In addition, core power maneuvering can be accomplished with control rods, soluble boron concentration changes, or a combination of control rods and soluble boron as described in Sections 4.2 and 4.3. Nuclear Steam Supply System The NSSS consists of a reactor core, two helical-coil SGs, and a pressurizer integrated within the RPV. The RPV is enclosed in an approximately cylindrical CNV that sits in the reactor pool. The reactor core is located below the helical-coil SGs inside the RPV. Using natural circulation, the primary reactor coolant flow path is upward through the central hot leg riser, and then downward around the outside of the SG tubes with return flow to the bottom of the core via an annular downcomer. As the reactor coolant flows across the SG tubes, heat is transferred to the secondary side fluid inside the SG tubes. Concurrently, as the secondary side 2 1.2-4 Revision 4.1

Reactor Core The core configuration for the NPM consists of 37 fuel assemblies and 16 control rod assemblies (CRAs). The CRAs are organized into two banks: a regulating bank and a shutdown bank. The regulating bank is used during normal plant operation to control reactivity. The shutdown bank is used during normal shutdown. All 16 CRAs are inserted for scram events. The fuel assembly design is similar to a standard 17x17 PWR fuel assembly with 24 guide tube locations for control rods and a central instrument tube. The only significant differences are the fuel assembly is nominally half the height of a standard fuel assembly and is supported by five spacer grids. The fuel is uranium dioxide, UO2, with gadolinium oxide, Gd2O3, as a burnable absorber homogeneously mixed within the fuel in select rod locations. The U-235 enrichment is less than 4.95 percent. A list of fuel design parameters is presented in Table 4.2-1. Pressurizer The pressurizer provides the primary means for controlling reactor coolant system (RCS) pressure. It is designed to maintain a stable reactor coolant pressure during operation. Reactor coolant pressure is increased by applying power to a pair of heater bundles installed above the pressurizer baffle plate. Pressure in the RCS is reduced using spray provided by the chemical and volume control system (CVCS). Steam Generator Each NPM uses two once-through, helical-coil SGs for steam production. The SGs are located in the annular space between the hot leg riser and the RPV inside diameter wall. The SG consists of tubes connected to feed and steam plenums with tube sheets. Preheated feedwater enters the lower feed plenum through nozzles on the RPV. As feedwater flows through the interior of the SG tubes, heat is transfered across the SG tube wall from the reactor coolant to the feedwater. The feedwater changes phase and exits the SG as superheated steam. Reactor Pressure Vessel The RPV consists of an approximately cylindrical steel vessel with an inside diameter of approximately 9 ft and an overall height of approximately 58 ft that is designed for an operating pressure of approximately 1,850 psia. The upper and lower heads are torispherical, and the lower portion of the vessel has a flange to provide access for refueling. The RPV consists of three sections: the RPV head section, the upper section, and the lower section. The RPV head is welded to the top of the upper section, and the upper and lower sections are flanged together using bolts. 2 1.2-5 Revision 4.1

reactor vent valves, reactor safety valves, reactor high point degasification instrumentation and controls (I&C) instrument channels, and the CRDM nozzles. The RPV upper section is cylindrical, approximately 9 ft in diameter with slightly thicker sections at the feedwater inlet and steam outlet areas. The upper section includes penetrations ranging from 2.25 to 25 diameter for main steam piping nozzles and main steam access ports, pressurizer heaters, feedwater piping nozzles and feedwater access ports, reactor recirculation valves, CVCS, and pressure instrumentation. The RPV lower section is cylindrical, approximately 9 ft in diameter and includes a torispherical lower head that is welded in place. There are no penetrations in the lower section of the RPV. A steel pressurizer baffle plate integral with the RPV provides a barrier between the saturated water in the pressurizer and the RCS. The pressurizer baffle plate is integrated with the upper steam plenums, has flow holes to allow surges of water into and out of the pressurizer, and to act as a thermal barrier. Containment Vessel The CNV is a cylindrical, steel pressure vessel housing the RPV, CRDMs, and associated NSSS piping and components. The CNV has an overall height of approximately 76 ft and an outside diameter of approximately 15 ft and consists of an upper CNV section with a welded torispherical top head and a lower CNV section with a welded head. The upper and lower CNV sections are flanged together using bolts. The flange connection permits the CNV to be separated to provide access to the RPV for refueling and maintenance. The safety functions of the CNV are to contain the release of radioactive material following postulated accidents and to provide heat rejection to the reactor pool following ECCS actuation. The CNV also provides support for the RPV. Manways provide access to components located inside the CNV. Penetrations on the CNV upper head are provided for process piping, electrical power, and instrumentation. The CNV is supported laterally by support lugs located slightly below the steam plenum elevation and by the support skirt attached to the CNV lower head. The support skirt also provides vertical support for the CNV. Internal to the CNV, the RPV is laterally and vertically supported by four support plates located slightly below the steam plenum elevation and is laterally supported at the center of the lower RPV head. The CNV is partially immersed in the reactor pool, which provides a passive heat sink for containment heat removal. The CNV is designed to withstand the external environment of the reactor pool as well as the internal pressure and temperature of a design-basis accident. 2 1.2-6 Revision 4.1

• minimizes moisture content that could impact the reliability and contribute to corrosion of components within the CNV
• facilitates detection of leakage from the reactor coolant pressure boundary
• eliminates convective heat transfer and therefore, the need for RPV insulation, which reduces potential debris generated in the CNV
• limits the initial amount of oxygen in containment (severe accident combustible gas consideration)

Following an actuation of the ECCS, steam is vented from the RPV through the reactor vent valves. This results in an initial spike in containment pressure and temperature. Steam in contact with the inside surface of the CNV is passively cooled and condensed by conduction and convection to the reactor pool water. This passive process rapidly reduces containment pressure and temperature and maintains containment pressure and temperature at less than design conditions indefinitely. 1.1.3 Safety Considerations NuScale has achieved an improvement in safety over existing plants through simplicity of design, reliance on passive safety systems, and small fuel inventory. The integral design of the NPM eliminates external coolant loop piping, which eliminates large-break LOCA scenarios. The availability of passive safety systems for decay heat removal, emergency core cooling, and control room habitability eliminates the need for external power under accident conditions. With these passive safety systems, small-break LOCAs do not challenge the safety of the plant. The result is a design with a core damage frequency that is lower than the current light water reactor fleet. The reactor core has a small radioactive source term as compared to a conventional 1,000 MWe nuclear reactor. Based on the smaller fuel inventory, the amount of radioactive material available for release during a postulated accident is reduced. Table 1.2-2 provides a listing of some of the features of the NPM. 1.2 Engineered Safety Features and Emergency Systems 1.2.1 Engineered Safety Feature Materials Details are provided in Section 6.1 related to the selection and fabrication methods for metallic and organic materials used in ESF components to ensure compatibility with fluids that the component may be exposed to during normal, accident, maintenance, and testing conditions. 1.2.2 Containment Systems The containment is an integral part of the NPM and provides primary containment for the RCS. Section 6.2 provides further information for the containment system. 2 1.2-7 Revision 4.1

The ECCS provides a passive means of decay heat removal in the event of a LOCA. The ECCS consists of three independent reactor vent valves and two independent reactor recirculation valves (Figure 1.2-9). All five valves are closed during normal operation. During ECCS operation, the reactor vent valves vent steam from the RPV into the CNV where the steam condenses and collects in the bottom of the containment. The reactor recirculation valves allow water to reenter the RPV and be circulated through the core. When reactor coolant temperature is reduced to below the boiling point, core cooling continues via conduction directly into the reactor pool. The cooling function of the ECCS is entirely passive, with heat being conducted through the CNV wall to the reactor pool. Section 6.3 provides design and operational information for the ECCS. 1.2.4 Control Room Habitability System The control room habitability system (CRHS) ensures that plant operators are adequately protected against the effects of accidental releases of toxic or radioactive gases. The CRHS is a passive system that provides clean, compressed, breathable air to the MCR in the event of a radioactive release or when AC power is not available. Areas served by the CRHS are maintained at positive pressure relative to adjacent areas. Compressed breathable air storage capacity can provide clean air to the MCR spaces for at least 72 hours following an initiating event. Section 6.4 provides design and operational information for the CRHS. 1.2.5 Fission Product Removal and Control Systems The only fission product removal and control system credited in the design is the CNV in conjunction with the containment isolation system. Fission product control is inherent in the design of the NPM, wherein the CNV atmosphere is depleted through the passive process of aerosol deposition. Section 6.5 provides information for this ESF. 1.2.6 Inservice Inspection of Class 2 and 3 Components The inservice inspection program includes the pre-service examinations and the periodic inservice inspections and tests necessary to ensure that safety-related and risk-significant systems, structures, and components are capable of fulfilling their intended safety functions. Section 6.6 provides detailed information for the inservice inspection program. 1.3 Instrumentation, Controls, and Electrical Systems The I&C architectural design philosophy incorporates clear interconnection interfaces, separation between safety and nonsafety systems, and simplification of system functions. The I&C architecture primarily consists of the following systems, which are described in Section 7.0: 2 1.2-8 Revision 4.1

• neutron monitoring system measures neutron flux as an indication of core power and provides safety inputs to the MPS.
• module control system (MCS) is a distributed control system that allows monitoring and control of module-specific plant components.
• plant control system supplies nonsafety inputs to the human system interfaces in the MCR, the remote shutdown station, and other locations where necessary.
• fixed area radiation monitoring system continuously monitors in-plant radiation and airborne radioactivity levels.
• safety display and indication system provides visual display and indication in the MCR from the MPS and plant protection system.
• plant protection system monitors and controls systems that are common to all NPMs and are not specific to an individual NPM.
• health physics network provides the permanently installed communications infrastructure necessary to support a licensee-implemented radiation protection program.
• in-core instrumentation system monitors various parameters within the reactor core and RCS and sends the parameter values to the MCS for display and evaluation.

Under normal operating conditions the AC electrical power distribution system supplies continuous power to equipment required for startup, normal operation, and shutdown of the plant. As described in Section 8.3, the NuScale Power Plant does not require onsite or offsite AC electrical power to cope with design-basis events. Safety systems are not reliant on AC or DC electrical power for actuation. The power systems within the plant are described below:

• The 13.8 KV and switchyard system provides power from the turbine generators and the auxiliary AC power source to the 13.8 kV AC buses and connects the onsite AC system to the switchyard.
• Medium voltage AC electrical distribution system provides power at 4,160V AC to buses servicing medium voltage loads.
• Low voltage AC electrical distribution system provides power at 120V AC and 480V AC to buses servicing low voltage loads.
• Highly reliable DC power system provides a failure-tolerant source of 125V DC power to plant loads including emergency lighting, MPS, PPS, and post-accident monitoring loads.
• Normal DC Power System provides power to nonsafety control and instrumentation loads.
• Backup power is provided for onsite AC power. The backup diesel generators provide power at the 480VAC level and the auxiliary AC power source provides power at the 13.8kVAC level.

2 1.2-9 Revision 4.1

The power conversion systems associated with an NPM consist of a main steam system, a turbine generator set, a standard condenser and cooling tower arrangement, and a condensate and feedwater system as shown in Figure 1.2-3. With multiple NPMs per plant, individual NPMs can be placed into service incrementally to meet construction schedules and grid demand as permitted by the site license. NPMs can also be taken off-line individually for refueling outages and maintenance. 1.5 Fuel Handling and Storage Systems The fuel handling and reactor maintenance areas are located in the west end of the RXB and include space for the SFP, refueling pool, and dry dock. The pools are shown in Figure 1.2-16. 1.6 Plant Cooling Water Systems The plant cooling water systems include several systems that are important to supporting plant operation. These systems include the following:

• The reactor component cooling water system (RCCWS) is a nonsafety-related, closed-loop cooling system that transfers heat from various plant components to the site cooling water system. The RCCWS provides cooling to the CRDMs, the non-regenerative heat exchangers for each CVCS, and the primary sampling system coolers. (Section 9.2.2)
• The reactor pool cooling system and the spent fuel pool cooling system are nonsafety-related, closed-loop systems that transfer heat from the associated pool to the site cooling water system. (Section 9.1.3)
• The circulating water system is an open-loop system that provides a continuous supply of cooling water to the plant turbine condensers. Circulating water pumps draw water from a common basin to provide cooling water flow for up to six condensers in one TGB. Heated circulating water from the outlet of the condensers flows to a set of mechanical-draft cooling towers where excess heat is removed as the water gravity flows back to the common basin. (Section 10.4.5)
• The site cooling water system is an open-loop system that provides a continuous supply of cooling water to the chilled water system, the balance of plant component cooling water system, the spent fuel pool cooling system, the reactor pool cooling system, the RCCWS, and the condenser air removal system. Site cooling water pumps draw water from a common basin to provide cooling water flow to the systems serviced. Heated site cooling water from the outlet of the individual system heat exchangers continues to a dedicated set of mechanical-draft cooling towers where excess heat is removed as the water gravity flows back to the common basin. (Section 9.2.7) 2 1.2-10 Revision 4.1

The radioactive waste management system is discussed in detail in Chapter 11. Liquid, gaseous, and solid radioactive waste management systems are discussed in detail in Sections 11.2, 11.3, and 11.4, respectively. Process effluent radiation monitoring and sampling systems are discussed in Section 11.5. 2 General Arrangement of Major Structures and Equipment Figure 1.2-2 presents the layout of a NuScale Power Plant. This figure includes an administration and training building and a warehouse that are outside the scope of the FSAR and not discussed further. 2.1 Reactor Building As shown in Figure 1.2-2, the RXB is approximately central to the site. See Figure 1.2-5 and Figure 1.2-10 through Figure 1.2-20 for RXB drawings. Dimensions provided in Figure 1.2-5 are nominal or approximate values for illustrative purposes. The RXB houses the NPMs and systems and components required for plant operation and shutdown. The RXB is primarily a rectangular configuration that is approximately 350 ft long and 150 ft wide, and extends approximately 81 ft above nominal plant grade level. The bottom of the RXB foundation is 86 ft below grade except for the areas under the elevator pit and the refueling pool, which are approximately 92 ft below grade. The RXB is a Seismic Category I, reinforced concrete structure with design considerations for the effects of aircraft impact, environmental conditions, postulated design basis accidents (internal and external), and design basis threats. The RXB also provides radiation protection to plant operations and maintenance personnel. Each NPM is located in the common reactor pool in its own three-walled bay with the open wall towards the center of the pool. The bays are arranged into two rows with six bays per row along the north and south walls of the reactor pool at the east end of the pool. A central channel is provided between the bays to allow for movement of the NPMs between the bays and the refueling pool. The bays are approximately 20 ft wide by 20 ft long by 98 ft deep with a normal reactor pool water depth of approximately 69 ft (this correlates to an elevation of approximately 94'). Each bay has a concrete bioshield to reduce radiation levels in the RXB and to prevent deposition of foreign materials onto an NPM. The bioshield consists of a two foot thick horizontal slab comprised of reinforced concrete with a stainless steel surface and a vertical assembly comprised of a square stainless steel tube framing system and series of radiation panel assemblies that extends into the pool. The horizontal slab is bolted to the top of the bay. Nine radiation panels are attached on both sides of the vertical bioshield framing system to provide a radiation barrier and ventilation. The bioshields are designed to be removed to access the NPM. To accommodate the removed bioshield, each bioshield is designed to have another bioshield stacked on top of it to allow for NPM movement during refueling. The NPM, reactor pool, and SFP are below grade. The surface of the reactor pool water is approximately 6 feet below grade. Also located below grade are most primary systems and some radioactive waste equipment. Hoisting and handling equipment is located above grade. 2 1.2-11 Revision 4.1

outages, and during replacement or removal of NPMs. Table 3.2-1, Classification of Structures, Systems, and Components, provides the location and classification of systems, structures, and components. 2.1.1 Fuel Handling and Reactor Maintenance Areas The fuel handling and reactor maintenance areas are located in the west end of the RXB and include space for the SFP, refueling pool, and dry dock. The pools are shown in Figure 1.2-16. The operating areas at the west end, 100'-0" elevation of the RXB provide space for the operation of fuel handling equipment and accessing the upper portion of an NPM while the reactor core is being refueled. The refueling pool is connected directly to the reactor pool accommodating transport of an NPM through the pool water using the Reactor Building crane (RBC). A weir between the refueling pool and SFP provides access for fuel assembly transport under water during the refueling process. The fuel handling and maintenance areas are designed to provide radiation protection for plant operations and maintenance personnel who are working in those areas. The area west of the SFP contains a fuel receiving area and a jib crane for loading new fuel assemblies into the new fuel elevator. The area has pallet jack access to aid in new fuel receiving activities. Upon receipt, new fuel assemblies are inspected and temporarily stored in racks in the SFP before being placed in a reactor core. The SFP provides storage space for the accumulated spent fuel assemblies prior to removal for dry storage and for temporary short-term storage for new fuel assemblies. Spent fuel assemblies removed from the reactor core are placed in spent fuel storage racks in the SFP. The refueling pool contains the bolting tools to disassemble and reassemble the NPM during refueling. The reactor core remains in the lower head of the RPV while in the refueling pool for refueling and fuel management. A fuel handling machine moves new and used fuel through the weir between the refueling pool and SFP. The dry dock area contains the module inspection rack and is separated from the refueling pool by a gate. With the gate closed, the dry dock water level can be lowered and maintenance activities on the upper NPM can be completed. Necessary inspection and testing equipment for the NPM are moved to this area during refueling. The dry dock provides maintenance access to the upper section of the NPM. The dry dock is also used for placing new NPM components into the reactor pool and preparing them for assembly. Additionally, it provides access for shipment of used NPMs off-site. 2 1.2-12 Revision 4.1

Refueling operations for an individual NPM is independent of the operating status of the remaining NPMs. During refueling, an NPM is moved from its operating bay in the reactor pool to the refueling pool using the RBC. The RBC lifts the NPM off its supports within the reactor bay and moves it to the open channel in the center of the reactor pool, which serves as a pathway to transport the NPM to the refueling area. In the refueling area, the NPM is set into the containment flange tool where the CNV flange is unbolted. The crane lifts the NPM, separating the lower CNV from the upper CNV with RPV still attached and intact. Next, the crane moves the upper CNV and RPV to the reactor vessel flange tool where the RPV flange is unbolted. The crane again lifts the NPM, this time separating the upper and lower RPV, leaving the lower RPV including the reactor core, in the reactor vessel flange tool. Finally, the crane transports the upper NPM (now consisting of just the upper CNV with attached upper RPV) to the module inspection rack in the dry dock. Inspection, testing, and maintenance are performed while the core is being refueled using a dedicated fuel handling machine. After inspection, maintenance, and testing are complete and the reactor core has been refueled, the upper portion of the NPM is moved from the dry dock to the refueling pool where the NPM is reassembled in reverse order using the dedicated flange tools. Following reassembly, the NPM is moved into the reactor pool and returned to its operating bay by the RBC. In the operating bay, startup tests are performed and the reactor is prepared for restart. After the NPM has passed necessary tests and inspections, and the reactor coolant is at startup conditions, the NPM is brought online, and steam and power production begins. 2.2 Control Building The CRB is located approximately 30 ft east of the RXB. See Figure 1.2-21 through Figure 1.2-27 for CRB drawings. The overall CRB footprint is rectangular, approximately 120 ft long by 80 ft wide at the 100'-0" elevation. The following portions of the CRB are nonsafety-related and Seismic Category II:

• above the 120'-0" elevation
• inside the elevator shaft (full building height)
• inside the two stairwells (full building height)
• the fire protection vestibule located on the East side of the CRB Structural steel and metal siding are used above the 120'-0" elevation. The remaining portion of the CRB, below the 120'-0" elevation, is a safety-related, Seismic Category I, concrete structure.

The lowest elevation of the CRB primarily houses electrical equipment and CRHS air bottles. There is a tunnel that connects the RXB to the CRB. The tunnel has two levels; 2 1.2-13 Revision 4.1

The MCR and the associated spaces are located below grade in the CRB. This is the area serviced by the CRHS. Associated spaces for the MCR include the following:

• conference room (shift turnover)
• open office area (auxiliary operator room)
• two offices
• storage room
• janitor closet
• three air locks
• viewing area
• shift managers office
• reference room
• emergency equipment room
• lavatories
• break room
• telecommunication room The technical support center (TSC) and the associated spaces are located at grade level in the CRB. Associated spaces for the TSC include:
• records storage
• three offices
• two conference rooms
• data equipment room
• lavatories
• data maintenance room
• break room Additional equipment located in the CRB includes the control room HVAC system (CRVS) equipment, the chilled water system equipment supporting the CRVS, and an elevator machine room.

2.2.1 Main Control Room The MCR contains control panels for all installed NPMs. Each reactor operator monitors and controls multiple NPMs from a control room panel. Figure 18.7-1 provides the layout for the MCR. 2 1.2-14 Revision 4.1

Each reactor control system display provides the monitoring for a specific reactor. Additional display stations, including a separate display for shared plant systems, provide control room operators with access to a wide range of plant information for trending and diagnostics. The reactor operators monitor the automated control system for each NPM. The MCR contains all alarms, displays, and controls for effective monitoring and control by the operators. The control room supervisor station provides an overview of all NPMs using multiple monitors. All monitor displays are designed using human factors analysis to enhance simplicity. The display layout and design uses graphical representations of plant systems and components. The following monitoring and control activities are typical control room functions:

• initiate NPM startup
• initiate NPM shutdown
• set or correct selected set points that control the NPM or plant functions
• take corrective actions if an NPM or plant system does not operate as intended The MCR enhances supervisory control of the NPMs and plant systems by providing alarm annunciation on the plant group-view overview display monitor as part of the alarm management system. This system includes information from the individual NPMs via the MPS, the MCS, and the shared I&C systems common to all the NPMs.

2.2.2 Technical Support Center A TSC is provided, compliant with the design requirements of NUREG-0696. Section 13.3 provides additional information. 2.3 Radioactive Waste Building The RWB houses equipment and systems for processing radioactive gaseous, liquid, and solid waste and for preparing waste for off-site shipment. See Figure 1.2-28 through Figure 1.2-33 for RWB drawings. The building houses equipment to prepare low-level radioactive waste for compaction to reduce volume and provides temporary storage for radioactive waste. HVAC equipment for high-efficiency particulate air filtration of air from the RXB and RWB is located in the RWB. The building is designed to maintain radiation exposures to operators and maintenance personnel as low as reasonably achievable. 2.4 Major Systems 2.4.1 Decay Heat Removal System The DHRS provides secondary side reactor cooling for non-LOCA events when normal feedwater is not available. The system, as shown in Figure 1.2-8, is a 2 1.2-15 Revision 4.1

capable of removing 100 percent of the decay heat load and cooling the RCS. Each train has a passive condenser immersed in the reactor pool. In the event of a SG tube failure, the affected SG is isolated and the DHRS provides cooling through the intact SG. On receipt of an actuation signal, feedwater and main steam isolation valves are closed and the DHRS valves open. Reactor coolant continues to circulate through the RPV collecting decay heat from the core. As water from the DHRS condenser travels through the SG tubes it is converted to steam absorbing decay heat from the reactor coolant. The steam then flows back to the DHRS condenser where it gives up excess heat to the reactor pool water and is condensed, and the cycle is repeated. This transfer of heat promotes natural circulation in both the RCS and the DHRS. Section 5.4.2 provides design and operational information for the DHRS. 2.4.2 Ultimate Heat Sink The ultimate heat sink is a large, stainless steel-lined, reinforced concrete pool located in the RXB below plant grade level. The ultimate heat sink consists of the reactor pool area, the refueling pool area, and the spent fuel pool area. The pool areas are shown in Figure 1.2-16. During normal plant operations, heat is removed from the pool through the reactor pool cooling system and rejected into the atmosphere through a cooling tower or other external heat sink. The spent fuel pool has an independent spent fuel pool cooling system. In a design basis accident involving a sustained loss of all AC power, decay heat is removed from the NPMs through passive heat transfer to the pool resulting in pool heat up and boiling. Water inventory in the reactor pool is adequate to cool the NPMs for at least 72 hours without adding water. The reactor pool provides an additional means of fission product retention beyond that of the fuel, fuel cladding, RPV, and the containment for certain events. Section 9.2.5 provides design and operational information for the ultimate heat sink. 2.4.3 Chemical and Volume Control System The CVCS is simple in design and its operation is not credited during or after an accident. During normal operation, the CVCS recirculates a portion of the reactor coolant through demineralizers and filters to maintain reactor coolant cleanliness and chemistry. A portion of the recirculated coolant is used to supply pressurizer spray for controlling reactor pressure. Reactor coolant inventory is controlled by injection of additional water when reactor coolant levels are low or letdown of reactor coolant to the liquid radioactive waste system when coolant inventory is high. 2 1.2-16 Revision 4.1

circulation flow in the RCS. Boron concentration in the RCS is controlled by a feed-and-bleed process. Injection pumps provide borated water or clean demineralized water that is delivered into the RCS with excess reactor coolant being letdown to the radioactive waste system. Safety-related protection is provided for an anticipated operational occurrence involving unintended dilution of the RCS due to CVCS equipment failure or operating error. Section 9.3.4 provides design and operational information of the CVCS. 2.5 Other Site Structures 2.5.1 Turbine Generator Building A NuScale Power Plant has two separate TGBs. The TGBs are nonsafety-related structures. Each building houses six turbine generator sets along with their auxiliaries, the condensers, condensate systems, and the feedwater systems. A laydown area and overhead crane are provided for installation and maintenance activities in each TGB. 2.5.2 Annex Building The ANB is a nonsafety-related structure. The ANB houses several facilities and serves several functions, including:

• controlling access to both radiologically-controlled and nonradiologically-controlled areas of the RXB
• housing various personnel support services such as locker rooms, showers, toilet facilities, lunch and conference rooms, and first aid

        *   ((providing space for personnel and component decontamination equipment and employee dosimeter processing

• housing a portion of the facilities that support plant security such as secondary alarm station, security briefing room, armory, security manager's office, etc.))

2.5.3 Security Buildings The SCBs are nonsafety related structures that include the following structures:

• primary access control building
• main security building
• vehicle barrier system The SCBs provide the following nonsafety-related functions:
• control personnel and vehicle entry into the PA and screen personnel seeking unescorted access into the PA.

2 1.2-17 Revision 4.1

• provide a structure or space to monitor access into areas of the plant as well as monitoring tamper alarm devices.

2.5.4 Central Utility Building The CUB is a nonsafety-related structure that houses common utility plant services, which include the following:

             *   ((chiller equipment

• instrument air system
• service air system
• chemical treatment equipment for demineralized water
• maintenance area
• life safety
• demineralized water equipment
• security functions))

2.5.5 ((Diesel Generator Buildings The NuScale Power Plant design includes two DGBs, each housing a single backup diesel generator. The principal functions of each DGB are to provide support and housing for the backup diesel generators and their auxiliary equipment. The DGB houses no safety-related systems and has no functional requirements that support the ESFs. The DGBs house the following:

• diesel engines and associated support equipment
• generators
• DGB HVAC system
• maintenance area))

3 Plant Features of Special Interest Human Factors Considerations The NuScale Power Plant design minimizes human error through fail-safe design functionality, allows multi-modular control capability from a single control room with effective automation design, employs digital display design and soft control technology to enhance usability, and provides optimum workload management. The HFE program satisfies specific regulatory requirements and guidance, and leverages human performance and operating experience from nuclear and non-nuclear industries. Chapter 18 describes the HFE Program. 2 1.2-18 Revision 4.1

Overall Plant inal net output 570 MWe* ber of power modules 12 Power Module ber of reactors One mal power rating 160 MWth inal gross electrical output 50 MWe normal operating pressure 1,850 psia m generator number Two m generator type Vertical helical tube m cycle Rankine-subcritical regenerative with superheat ine type 3,600 rpm, condensing, with extraction Reactor Core UO2 (<4.95% enrichment) eling intervals 24 months minal net output is total gross electrical output minus house loads. 2 1.2-19 Revision 4.1

NuScale Design Feature Primary Impact Safety Enhancement contained within the RPV No large diameter primary coolant Eliminates postulated large-break LOCA piping spectrum accidents ral-convection-cooled core No reactor coolant pumps Eliminates reactor coolant pump accidents, shaft breaks, pump seizure, missile generation and pump leaks containment design pressure Containment peak pressure for Containment integrity assured, minimizing the worst case design-basis accident potential for radioactive releases during remains below containment design postulated accidents. pressure and NSSS inside the CNV During an accident, any water lost No postulated design-basis small-break LOCA from RPV stays within containment capable of uncovering nuclear fuel and is returned to the RPV by passive means uated containment Subatmospheric pressure during Minimal amount of noncondensible gases normal operation increases the steam condensation rate for containment heat removal during postulated small-break LOCA. Amount of oxygen in containment during normal operations is minimized. No insulation on RPV Eliminates potential sump screen blockage and permits cooling of the exterior of the vessel during an accident power core (160 MWt) Reduces decay heat removal Enhances in-vessel retention; maintains low requirements accident consequences; reduces fission product source term; simplifies emergency planning tor pool with partially CNV partially immersed in reactor Provides passive long-term cooling roximately 90%) immersed NPM pool ive safety systems Safety systems cool and Active safety systems are not required depressurize the RPV/CNV even in the event of loss of external power 2 1.2-20 Revision 4.1

cale Final Safety Analysis Report General Plant Description cale Final Safety Analysis Report Owner Controlled Area NUSCALE POWER PLANT Security Owner Controlled Area (SOCA) Fence Cooling Tower Protected Area (PA) Boundary Double Fence BDG ANB TGB Admin/Training Structures within the CERTIFIED DESIGN NUSCALE POWER MODULES (NPMS) AAP SEC SEC RWB RXB CRB CUB SWYD Warehouse SCWS Pumps TGB SCWS Tower BDG Protected Area (PA) Boundary Double Fence Cooling Tower Security Owner Controlled Area (SOCA) Fence General Plant Description Certified Design Structures or Components NuScale Power Modules Structures not described in the FSAR Conceptual Design Structures containing Certified Design SSCs Owner Controlled Area boundary Conceptual Design SSCs NuScale Power Plant boundary (SOCA Fence)

cale Final Safety Analysis Report General Plant Description Tier 2 NuScale Final Safety Analysis Report Figure 1.2-4: Layout of a Multi-Module NuScale Power Plant (( Withheld - See Part 9 }} 1.2-24 General Plant Description Revision 4.1

Tier 2 NuScale Final Safety Analysis Report Figure 1.2-5: Cutaway Illustration of 12 Module Configuration (( Withheld - See Part 9 }} 1.2-25 General Plant Description Revision 4.1

2 1.2-26 Revision 4.1 2 1.2-27 Revision 4.1 2 1.2-28 Revision 4.1 2 1.2-29 Revision 4.1 NuScale Final Safety Analysis Report General Plant Description Figure 1.2-10: Reactor Building 24-0 Elevation (( Withheld - See Part 9 }} Tier 2 1.2-30 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-11: Reactor Building 35'-8" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-31 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-12: Reactor Building 50'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-32 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-13: Reactor Building 62'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-33 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-14: Reactor Building 75'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-34 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-15: Reactor Building 86'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-35 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-16: Reactor Building 100'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-36 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-17: Reactor Building 126'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-37 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-18: Reactor Building 145'-6" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-38 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-19: Reactor Building East and West Section View (( Withheld - See Part 9 }} Tier 2 1.2-39 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-20: Reactor Building South Section View (( Withheld - See Part 9 }} Tier 2 1.2-40 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-21: Control Building 50'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-41 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-22: Control Building 63'-3" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-42 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-23: Control Building 76'-6" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-43 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-24: Control Building 100'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-44 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-25: Control Building 120'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-45 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-26: Control Building North Section View (( Withheld - See Part 9 }} Tier 2 1.2-46 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-27: Control Building West Section View (( Withheld - See Part 9 }} Tier 2 1.2-47 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-28: Radioactive Waste Building 71'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-48 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-29: Radioactive Waste Building 82'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-49 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-30: Radioactive Waste Building 100'-0" Elevation (( Withheld - See Part 9 }} Tier 2 1.2-50 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-31: Radioactive Waste Building 120-0 Elevation (( Withheld - See Part 9 }} Tier 2 1.2-51 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-32: Radioactive Waste Building North and South Section Views (( Withheld - See Part 9 }} Tier 2 1.2-52 Revision 4.1

NuScale Final Safety Analysis Report General Plant Description Figure 1.2-33: Radioactive Waste Building West Section View (( Withheld - See Part 9 }} Tier 2 1.2-53 Revision 4.1

The major NuScale Power Plant design features and nominal parameters are provided in Table 1.3-1 and discussed further in the associated final safety analysis report (FSAR) section(s). These NuScale features and values are shown in comparison with a typical pressurized water reactor (PWR) plant design. All values are nominal and provided for comparison only. The typical PWR values presented are representative of the Standardized Nuclear Unit Power Plant System design. Table 1.3-2 provides a comparison of safety systems and components required to protect the reactor core for the NuScale Power Plant versus a typical PWR plant. 2 1.3-1 Revision 4.1

cale Plant Parameter or Feature (per NPM) Typical PWR NuScale inal gross electrical output (MWe) 1,186 50 thermal output (MWt) 3,411 160 ber of fuel assemblies 193 37 assembly lattice -17x17 17x17 tive fuel length (ft) 12 6.56 rods per fuel assembly 264 264 age linear heat rate (kW/ft) 5.4 2.5 ber of Control Rod Assemblies 53 16 gn life (years) 40 60 tor Coolant System ber of heat transfer loops 4 No External Loops tor Coolant Pipes (in.) 27.5-31 None rating pressure (psia) 2,250 1,850 eg temperature (°F) 618 590 tor Vessel el inner diameter (in.) 173 107.5 mal shielding- and reflector design Neutron pad design Stacked stainless steel reflector blocks re instrumentation Bottom mounted Top mounted m Generator ber 4 2 Vertical U-tube Helical coil transfer area (ft2) 55,000 Approximately 18,000 ber of tubes 5,626 1,380 tor Coolant Pumps 4 0 surizer nal volume (ft3) 1,800 568 e nozzle nominal diameter (in.) 14 None dual Heat Removal Pumps 2 None ainment PCCV Steel Pressure Vessel r diameter (ft-in.) 140-0 14-2 ht (ft-in.) 205-0 (inner) 75-8.5 (outer) ainment Spray Pumps 2 None Pressure Safety Injection Pumps 2 None ging / Safety Injection Pumps 2 None Pressure Safety Injection Pumps 2 None mulators 4 None ystem type Analog Digital rgency Diesel Generators 2 None ine Type 1800 rpm, Tandem Compound Six 3,600 rpm, 10 stage with Superheat Flow rgency Feedwater Pumps 3 None ging Pumps (CVCS pumps) 2 2 for Safety Injection Yes No me Control Tank 1 0 tor Component Cooling Water Pumps 4 6 total for 12 NPMs 2 1.3-2 Revision 4.1

NuScale Comparison with Other Facilities Safety System or Component Typical PWR NuScale tor Pressure Vessel X X ainment Vessel X X tor Coolant System X X y Heat Removal System X X rgency Core Cooling System X X rol Rod Drive System X X ainment Isolation System X X ate Heat Sink X X dual Heat Removal System X ty Injection System X eling Water Storage Tank X densate Storage Tank X liary Feedwater System X rgency Service Water System X ogen Recombiner or Ignition System X ainment Spray System X tor Coolant Pumps X ty-Related Electrical Distribution System X native Off-Site Power X rgency Diesel Generators X ty-Related Class 1E Battery System X cipated Transient Without Scram (ATWS) System X 2 1.3-3 Revision 4.1

1 Not Used 2 Division of Responsibility NuScale Power, LLC (NuScale) has the overall design responsibility for the NuScale certified design. 3 Principal Consultants and Other Participants Fluor Corporation (Fluor) provided the balance of plant design described in the Design Certification Application. Item 1.4-1: A COL applicant that references the NuScale Power Plant design certification will identify the prime agents or contractors for the construction and operation of the nuclear power plant. 2 1.4-1 Revision 4.1

This section describes the verification and confirmation tests of unique design features that support the safety analysis for the NuScale Power Plant. The testing program described in this section was developed to provide data to support the final safety analyses. 1 NuScale Testing Programs The following testing programs have been completed or are currently in progress. The tests focus on design features of the NuScale Power Module (NPM) for which applicable data or operational experience did not previously exist. Tests specific to the NuScale fuel design are summarized in Section 1.5.1.1 and Section 1.5.1.2; tests specific to the steam generator (SG) are summarized in Section 1.5.1.3 and Section 1.5.1.4; tests specific to the control rod assemblies are summarized in Section 1.5.1.6 and Section 1.5.1.7; and tests involving integrated system phenomena are summarized in Section 1.5.1.5. 1.1 Critical Heat Flux Testing - Preliminary Fuel Design The NPM employs a fuel design for heat generation that is similar to a standard pressurized water reactor (PWR), with the exception of the fuel assembly height and the reactor coolant driving force. The NuScale fuel is approximately half the height of standard PWR fuel and features low-flow natural circulation of primary coolant rather than pump-driven primary coolant flow. In order to meet fuel licensing requirements, two critical heat flux (CHF) test programs were conducted: (1) a test program described in this section for the preliminary fuel design, and (2) a second test program described in Section 1.5.1.2 for the final fuel design. The preliminary fuel design test program supported code development and safety analysis efforts, and provided data for development of NuScale's NSP2 CHF correlation for the final fuel design. The NPM reactor core design employs 37 nuclear fuel assemblies. Each assembly is composed of a 17x17 square lattice of fuel rods assembled according to a given rod-to-rod pitch. Each fuel rod is approximately 2 meters in length. Fuel rods are assembled using spacer grids placed at specified locations along the length of the fuel rods such that fuel rods are evenly spaced and adequately supported. Primary coolant enters the NPM reactor core from the bottom through the core inlet plenum and heat transfer to the coolant occurs as coolant travels upward along the length of the fuel assemblies. In off-normal conditions, such as anticipated operational occurrences and postulated accidents, it must be known how close the heat transfer mode is to transitioning to a state where a continuous steam layer covers the fuel rods or portions of the fuel rods. The point at which this transition occurs is referred to as the CHF point. In order to determine the CHF point for the reduced-length fuel under appropriate flow conditions, a CHF testing program was conducted over a wide range of operating conditions. In these tests, instrumentation was used to measure key test parameters, including: resistance temperature detectors (RTD), thermocouples, pressure transducers, mass flow rate instruments, and electrical voltage and current meters. These sensors were used to measure heater rod temperatures and fluid flow conditions at various points of the fluid loop, and the electrical power supplied to heater rods when CHF occurred. The tests allowed NuScale to obtain fuel bundle subchannel exit 2 1.5-1 Revision 4.1

powers and hydraulic conditions. All information necessary for CHF correlation development and evaluation was collected. Testing for the preliminary fuel design CHF test series was performed at Stern Laboratories in Ontario, Canada using their existing high-pressure flow loop, electrical power supplies, heat rejection devices, water conditioning system, and data acquisition system. The test section and fuel simulators were designed and manufactured to represent the preliminary NuScale fuel design. Fuel assembly simulators with different axial power shapes were manufactured along with two sets of spacer grids that allowed testing of a 5x5 fuel assembly simulation bundle with or without the center fuel rod replaced by a guide tube. The approach of using a 5x5 representation of the larger 17x17 fuel assembly is an industry-accepted practice for CHF testing of PWR fuel bundles. The fuel assembly simulation bundles were mounted within a square flow channel and installed in the instrumented test section. This allowed testing of three separate configurations of fuel assembly simulation bundles with different axial flux shapes and fuel assembly subchannels as described in NuScale Power Critical Heat Flux Correlations topical report (TR-0116-21012). Tests were performed to envelope a range of bounding conditions and axial power shapes for vertical 5x5 fuel assembly configurations in accordance with the test specification and program documentation, which provided detailed test matrices for steady-state and transient CHF testing, pressure drop, and thermal mixing. The vertical 5x5 fuel assembly configurations were tested using industry-accepted test and acceptance methodology. The CHF testing was conducted by flowing water over the test sections at discrete test points covering a range of hydraulic conditions sufficient to develop a CHF correlation that spanned the NPM operational envelope. At each test point, the loop was configured for the specified flow, inlet temperature, and exit pressure conditions. The bundle power was increased until CHF was detected, which was indicated by an excursion of the fuel simulator thermocouples. Loop flow conditions (temperature, pressure, and flow), bundle power, rod power, and fuel simulator temperatures were recorded for each run. As-built data for the test section and test article, such as flow channel width, fuel simulator diameters, and spacer grid dimensions, were also recorded. In conclusion, tests were performed for a variety of thermal conditions using representative 5x5 fuel assembly simulations with a 2-meter heated length, differing axial power profiles, with and without a simulated guide tube. The testing investigated the effects of shorter fuel length and low-flow natural circulation of the primary coolant, and provided data that were used to develop NuScale's NSP2 CHF correlation in support of the NuScale small modular reactor technology. This test program was inspected by the NRC in accordance with Inspection Procedures IP 35034, 35017, and 36100. 2 1.5-2 Revision 4.1

The primary objective for this test program was to obtain CHF data for the NuScale fuel design that employs AREVA HMP'/HTP' spacer grid technology (designated as NuFuel HTP2') to augment the existing database that was previously obtained for NuScale's preliminary fuel design (described in Section 1.5.1.1). The new data was used to develop NuScale NSP4 CHF correlation and to validate NuScale's NSP2 CHF correlation developed using the preliminary fuel design tests for the NPM application. In addition, this test allowed NuScale to obtain bundle subchannel exit temperatures to determine mixing coefficients, and to collect single-phase and two-phase pressure drop characteristics of the assembly for a range of bundle powers and hydraulic conditions. The CHF test employed an electrically-heated test section that consisted of a 5x5 simulated fuel bundle built to prototypic geometry and employing AREVA HTP'/HMP' grid technology. The fuel assembly simulators with different power shapes were tested using a 5x5 fuel bundle with or without the center fuel rod replaced by a guide tube. The testing was conducted by flowing water through the test section at specified flow rates over a range of hydraulic conditions of the NPM. At each test point, the loop was configured for a specified flow rate, inlet temperature, and exit pressure conditions, and the bundle power was increased until CHF was detected over a range of operating conditions and axial power shapes for vertical 5X5 fuel assembly configurations. The occurrence of CHF was indicated by an excursion of the fuel simulator temperatures. The prototypic fuel design tests were conducted at the AREVA Karlstein Thermal Hydraulics (KATHY) facility located in Karlstein, Germany. The test data was used to validate the applicability of NuScale's NSP2 CHF correlation and to develop the NSP4 correlation for the NuFuel HTP2' fuel design. 1.3 Steam Generator Thermal-Hydraulic Performance Testing - Electrically Heated Facility The NPM incorporates two collocated SGs housed within the reactor pressure vessel. The SGs provide heat transfer to and from the primary system for both normal and off-normal conditions. Through natural circulation, the reactor coolant system transfers the core power to the SG converting feed water into steam. Unlike current PWR designs, the reactor coolant flows around the outside of the SG tubes (primary side) and the feedwater and main steam flow through the inside of the tubes (secondary side). Because these design aspects of the helical SGs are different from those used in the nuclear fleet, operational experience is not available and large-scale experimental data were needed for validation of NuScale thermal-hydraulic systems and design computer codes, as well as determination of SG performance characteristics. The objective of this testing was to determine the secondary side (inside tube) thermal-hydraulic performance of individual helical tubes representative of those used in the NPM steam generator design. This required testing over a range of conditions representative of the operational envelope. Measurement data were required to evaluate the distribution of temperature and pressure on the inside of the tubes. 2 1.5-3 Revision 4.1

Heating was accomplished using Joule heating, wherein a known electrical current is passed through the tube walls to produce a constant heat flux boundary condition on the inside of the tubes. Three distinctive heating zones were employed to provide different heat fluxes for the subcooled, boiling, and superheat regions. Within each zone the heat flux was constant, which represents a simplification from the heat flux profile that results when fluid heating is employed, as would occur in an operating NuScale SG. This approach enabled tube wall heat flux to be controlled during testing and permitted better access to instrumentation on the outside of the tubing. The testing was performed at the Societ Informazioni Esperienze Termoidrauliche (SIET) test facility in Piacenza, Italy. Types of testing carried out included adiabatic testing, diabatic testing, transient testing, and density wave oscillation testing. Dynamic pressure measurements were recorded during test runs which supported development of power spectral density spectra that may be used to support evaluation of the potential for internal two-phase (boiling) pressure fluctuations to contribute to flow induced vibration of SG tubes. These data also were used to inform sizing of the SG inlet flow restrictors for stable secondary-side SG operation, to provide benchmarking for NuScale thermal-hydraulic design and systems computer codes, and to define steam outlet conditions as a function of primary heat generation and secondary side conditions. This test program was inspected by the NRC in accordance with Inspection Procedures IP 35034, 35017, and 36100. 1.4 Steam Generator Thermal-Hydraulic Performance Testing - Fluid-Heated Facility Subsequent to the SG tests described in Section 1.5.1.3 that used three electrically heated SG tubes, a second set of SG tests was conducted using a 252-tube bundle array that was fluid heated on the exterior of the tubes to more accurately represent primary side SG conditions. The test facility included a large pressure vessel, which was able to accommodate the tube bundle test section and allowed for testing at elevated pressures and temperatures. The test facility included heaters and pumps that provided a span of flow rates at a wide range of thermal-hydraulic conditions. The fluid-heated test focused on overall primary and secondary side performance, and consisted of a bank of 252 helical tubes, modeling five of the 21 helical coil columns, operated at near-prototypic primary- and secondary-flow conditions. Testing activities were conducted at SIET in Piacenza, Italy using their fluid-heated hydraulic loop. Types of testing carried out included: adiabatic, diabatic, transient, density wave oscillation, and fluid-elastic instability tests. Each type of test consisted of multiple test points covering a range of conditions to characterize the phenomena of interest at various combinations of primary-side and secondary-side pressures, temperatures, and flow rates. In these tests, thermocouples, pressure transducers, mass flow rate instruments, and strain gauges were used to collect temperature, pressure, flow rate, and vibration data at several locations on the primary and secondary sides of the SG. These data have been used to benchmark NuScale thermal-hydraulic design and systems computer codes, and to define steam outlet conditions as a function of primary-fluid heating and secondary-side conditions. 2 1.5-4 Revision 4.1

The purpose of the NuScale integral system test program was to generate thermal-hydraulic data for system characterization and safety code validation using a scaled representation of the NPM design. Tests have also informed safety methodology development. The NuScale Integral System Test Facility (NIST-1) allows NuScale to replicate the integrated thermal-hydraulic phenomenon occurring in the reactor coolant system, containment, safety systems, and reactor pool. Data collected provide system characterization data required for validation of safety-related software, NRELAP5 and PIM. The NRELAP5 code, which is based on a commercial version of RELAP5-3D, is a thermal-hydraulic analysis code developed at NuScale for use in the thermal-hydraulic design, safety analysis, and licensing of the NPM. The code incorporates models and correlations specific to the unique design features of the NPM. The PIM code is a NuScale-developed proprietary code used to assess the stability characteristics of the NPM during operation. The NIST-1 is a scaled representation of the NPM reactor, containment, and reactor pool systems. It is constructed of stainless steel and has a maximum operating pressure of 1650 psia (11.4 MPa) and temperature of 630 degrees F (605 degrees K). NIST-1 volumes, lengths, and areas are obtained by multiplying the respective NPM design volumes, lengths, active heat transfer areas, and flow areas by the applicable scale factors determined through a detailed scaling analysis. This process ensures the NIST-1 properly captures the important thermal-hydraulic phenomena and processes that would occur in the plant. A series of tests have been completed at the NIST-1, located on the Oregon State University campus in Corvallis, OR, in support of NuScale's Design Certification Application. These tests include:

• facility characterization tests used to develop the NRELAP5 model of the NIST-1.
• loss-of-coolant accident (LOCA) tests used to validate NRELAP5 for LOCA and containment analyses.
• flow-stability tests used to validate PIM for reactor stability analyses.
• non-LOCA (anticipated operational occurrence) tests used to validate NRELAP5 for non-LOCA analyses.
• long-term cooling tests used to validate NRELAP5 for long term cooling analyses.

Data obtained from the NIST-1 tests identified above have been used to successfully validate the NRELAP5 and PIM codes for LOCA and containment, non-LOCA, flow stability, and long term cooling applications. This test program was inspected by the NRC in accordance with Inspection Procedures IP 35034, 35017, and 36100. 1.6 Control Rod Drive Mechanism Proof Test The control rod drive mechanism for the NPM contains features that are not common in conventional control rod drive mechanisms: a remote disconnect mechanism and a long control rod drive shaft. A proof-of-concept testing program was conducted to 2 1.5-5 Revision 4.1

is described in Section 1.5.1.7. Testing was completed at the Curtiss Wright facilities in Cheswick, PA, for both remote connect and remote disconnect operation of the coils. The test setup included a functional drive rod assembly, a prototypic remote disconnect gripper coil, a prototypic remote disconnect gripper latch, a prototypic lift coil, and weights to simulate the control rod assembly (CRA) with a prototypic CRA hub socket. The remote disconnect mechanism was found to provide a reliable and repeatable method to engage and disengage the CRA within the reactor pressure vessel. This is consistent with the results of the remote operation, lift verification, and manual disengagement testing that was performed. The tests provided a demonstration of hardware performance, which has been extended to aid in the design of drive rod position detection circuitry. Information gained from this testing has been used as a development tool to improve the design and does not create a design basis for the final control rod drive mechanism. 1.7 Control Rod Assembly Drop and Control Rod Drive Shaft Alignment Test The NPM is designed with control rod drive shafts that are longer than conventional PWR designs and have the capability to be remotely disconnected. The control rod drive shafts are aligned using the following multiple-support features:

• control rod drive mechanism nozzles in the reactor vessel head
• integrated steam plenum
• five upper control rod drive shaft supports in the upper riser section
• a control rod drive shaft alignment cone located at the top of the CRA guide tube The design uses a CRA and fuel-assembly design similar to, but shorter than, traditional operating reactors. The arrangement of a shorter fuel assembly and CRA coupled to a longer control rod drive shaft creates a unique configuration of these components with no operational or testing experience. The CRA-drop and control rod drive shaft-alignment test program was developed to confirm the operability of this unique design.

Testing was completed at the AREVA Technical Center in Erlangen, Germany, and was configured as an ambient pressure and temperature test. The ambient test configuration used a full-length control rod drive shaft coupled with a NPM control rod assembly and fuel assembly, as well as the control rod drive shaft support structures and a CRA guide tube assembly. The CRA guide tube assembly and fuel assembly were immersed in the water under ambient conditions with no coolant flow. During the test, the coupled CRA and control rod drive shaft assembly was dropped using multiple configurations having variations in the alignment of the control rod drive shaft supports and CRA guide tube assembly and a mid-span deflection of the fuel assembly. 2 1.5-6 Revision 4.1

time and CRA impact force at end of drop. 1.8 Emergency Core Cooling System Valve Design Certification Application Demonstration Test The NPM design utilizes emergency core cooling system (ECCS) valves. The reactor vent valves (RVVs) are located on the reactor vessel pressurizer, and the reactor recirculation valves (RRVs) are located on the reactor vessel downcomer. The RVVs and RRVs are functionally similar in design, however the RVVs are larger than the RRVs. Each of the ECCS valves are pilot operated by remote-mounted trip and reset solenoid valves located at the reactor containment boundary. An inadvertent actuation block (IAB) feature is included in the system, located inside containment on each main valve. A test program was developed to demonstrate functional performance of the ECCS valve system design including the unique aspects of the design, such as the configuration of the ECCS valve components and the IAB feature. The purpose of this test program was to

• demonstrate the ECCS valves function at reactor operating pressures and temperatures.
• demonstrate the ECCS valves function in borated reactor coolant.

Testing activities were performed at the Curtiss-Wright Valve Group Target Rock facility in Farmingdale, New York, using high pressure and temperature valve test cells. A test article was designed to represent an RRV consisting of the main valve, trip valve, reset valve, and IAB valve components. To represent the NPM design, a pilot-operated main valve was used with an inlet connected to a source able to provide water at reactor operating pressures and temperatures, and an outlet that exhausted to atmosphere. Trip line tubing and a trip valve representative of the NPM design was connected to the main valve to simulate NPM valve actuation performance and potential for boric acid buildup in internal passages. A reset valve was used to pressurize and close the main valve prior to each test run. A representative IAB valve was used in the same functional configuration as the NPM design. Types of tests carried out included main valve actuation, IAB functionality, and boric acid effects. Boric acid effects testing was performed using a vessel simulating the main valve control chamber in place of the main valve. Boric acid concentrations were selected to bound refueling and operating boron concentrations for the NPM design. Tests were performed through the range of reactor operating pressures and temperatures. This test program was inspected by the NRC in accordance with Inspection Procedures IP35034. 2 NuScale Test and Inspection Plans Information on NuScale test and inspection plans related to plant startup testing is provided in Section 14.2. 2 1.5-7 Revision 4.1

Topical reports and technical reports that are incorporated by reference as part of the NuScale Power Plant Design Certification Application are listed in Table 1.6-1 and Table 1.6-2, respectively. 2 1.6-1 Revision 4.1

NuScale Final Safety Analysis Report Material Referenced Table 1.6-1: NuScale Referenced Topical Reports Topical Report Number Topical Report Title Submittal Date FSAR Section NP-TR-1010-859-NP, Rev 4 NuScale Topical Report: Quality Assurance June 2019 17 Program Description for the NuScale Power Plant TR-0515-13952-A, Rev 0 Risk Significance Determination July 2015 17, 19 TR-0815-16497-P-A, Rev 1 Safety Classification of Passive Nuclear Power February 2018 8, 15 Plant Electrical Systems TR-1015-18653-P-A, Rev 2 Design of the Highly Integrated Protection September 2017 7, 15 System Platform Topical Report TR-0915-17565-P-A, Rev 4 Accident Source Term Methodology February 2020 15 TR-0116-20825-P-A, Rev 1 Applicability of AREVA Fuel Methodology for February 2018 4 the NuScale Design TR-0616-48793-P-A, Rev 1 Nuclear Analysis Codes and Methods November 2018 4 Qualification TR-0516-49417, Rev 1 Evaluation Methodology for Stability Analysis of August 2019 4 the NuScale Power Module TR-0516-49422-P, Rev 1 LOCA Evaluation Model November 2019 15 TR-0915-17564-P-A, Rev 2 Subchannel Analysis Methodology February 2019 4 TR-0516-49416-P, Rev 2 Non-LOCA Methodologies November 2019 15 TR-0116-21012-P-A, Rev 1 NuScale Power Critical Heat Flux Correlations December 2018 4 TR-0716-50350-P, Rev 1 Rod Ejection Accident Methodology November 2019 15 TR-0716-50351, Rev 0 NuScale Applicability of AREVA Method for the September 2016 4 Evaluation of Fuel Assembly Structural Response to Externally Applied Forces Tier 2 1.6-2 Revision 4.1

cale Final Safety Analysis Report port Number Title FSAR Section 116-20781 Fluence Calculation Methodology and Results 4.3, 5.3 316-22048 Nuclear Steam Supply System Advanced Sensor Technical Report 7.1, 7.2 416-48929 NuScale Design of Physical Security Systems 9.5, 13.6, 14.2, 14.3 516-49084 Containment Analysis Methodology 6.2 616-49121 NuScale Instrument Setpoint Methodology Technical Report 7.0, 7.2 716-50424 Combustible Gas Control 3.8, 6.2 716-50439 Comprehensive Vibration Assessment Program (CVAP) Technical Report TR-0716-50439 3.9, 14.2 816-49833 Fuel Storage Rack Analysis 3.7, 3.8, 9.1 816-50796 Loss of Large Areas Due to Explosions and Fires Assessment 20.2 816-50797 Mitigation Strategies for Loss of All AC Power Event 20.1 816-51127 NuFuel HTP2 Fuel and Control Rod Assembly Designs 4.2 916-51299 Long-Term Cooling Methodology 5.4, 6.2, 6.3, 15.0 916-51502 NuScale Power Module Seismic Analysis 3.7, 3.12, 3B 015-18177 Pressure and Temperature Limits Methodology 5.3 016-51669 NuScale Power Module Short-Term Transient Analysis 3.9 116-51962 NuScale Containment Leakage Integrity Assurance 6.2 116-52065 Effluent Release Methodology Technical Report 11.1, 11.2, 11.3 215-10815 Concept of Operations 18.7 316-17614 Human Factors Engineering Operating Experience Review Results Summary Report 18.2 316-17615 Human Factors Engineering Functional Requirements Analysis and Function Allocation Results Summary Report 18.3 316-17616 Human Factors Engineering Task Analysis Results Summary Report 18.4 316-17617 Human Factors Engineering Staffing and Qualifications Results Summary Report 18.5 316-17618 Human Factors Engineering Treatment of Important Human Actions Results Summary Report 18.6 316-17619 Human Factors Engineering Human-System Interface Design Results Summary Report 18.7 516-49116 Control Room Staffing Plan Validation Results 18.5 914-8534 Human Factors Engineering Program management Plan 18.1 914-8543 Human Factors Verification and Validation Implementation Plan 18.1 914-8544 Human Factors Engineering Design Implementation Implementation Plan 18.11 215-20253 Control Room Staffing Plan Validation Methodology 18.5 Material Referenced 917-56119 CNV Ultimate Pressure Integrity 3.8 918-60894 Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report 3.9, 14.2 818-61384 Pipe Rupture Hazards Analysis 3.6 304-1381 Human-System Interface Style Guide 18.10 018-61289 HFE Verification and Validation Results Summary Report 18.1

Where appropriate, simplified instrumentation and controls (I&C), electrical, or mechanical drawings are provided as figures. These figures are used in conjunction with the written text in the associated section to provide visual clarification of design information. Component position indications shown on these figures do not represent a specific operational state unless noted. 1 Electrical and Instrumentation and Control Drawings Table 1.7-1 provides a list of I&C functional diagrams and electrical one-line diagrams used in the FSAR. See Figure 1.7-1a, Figure 1.7-1b, and Figure 1.7-2 for the legends of the symbols and characters used in electrical and I&C diagrams. Item 1.7-1: A COL applicant that references the NuScale Power Plant design certification will provide site-specific diagrams and legends, as applicable. 2 Piping and Instrumentation Diagrams Table 1.7-2 provides a list of system drawings used in the FSAR. See Figure 1.7-3a through Figure 1.7-3f for a legend of the symbols and characters used in piping and instrumentation diagrams. Item 1.7-2: A COL applicant that references the NuScale Power Plant design certification will list additional site-specific piping and instrumentation diagrams and legends as applicable. 2 1.7-1 Revision 4.1

re Title re 7.0-1 Overall Instrumentation and Controls System Architecture Diagram re 7.0-3 Module Protection System Safety Architecture Overview re 7.0-4 Separation Group A Communication Architecture re 7.0-5 Separation Group A and Division I Reactor Trip System and Engineered Safety Features Actuation System Communication Architecture re 7.0-6 Reactor Trip Breaker Arrangement re 7.0-7 Equipment Interface Module Configuration re 7.0-8 Equipment Interface Module Output re 7.0-9 Pressurizer Trip Breaker Arrangement re 7.0-10 Module Protection System Gateway Diagram res 7.0-11a and 7.0-11b Module Protection System Power Distribution re 7.0-12 Neutron Monitoring System Ex-Core Block Diagram re 7.0-13 Plant Protection System Block Diagram re 7.0-14 Safety Display and Indication System Boundary re 7.0-15 Safety Display and Indication Hub re 7.0-16 Display Interface Module re 7.0-17 Module Control System Internal Functions and External Interfaces re 7.0-20 Plant Control System Internal Functions and External Interfaces re 8.3-1 Station Single Line Diagram res 8.3-2a and 8.3-2b 13.8kV and Switchyard System res 8.3-3a and 8.3-3b Medium Voltage Alternating Current Electrical Distribution System res 8.3-4a through 8.3-4z Low Voltage Alternating Current Electrical Distribution System res 8.3-5a and 8.3-5b Backup Power Supply System re 8.3-6 Highly Reliable Direct Current Power System (Common) res 8.3-7a and 8.3-7b Highly Reliable Direct Current Power System (Module Specific) res 8.3-8a through 8.3-8f Normal Direct Current Power System re 11.5-2 Process and Effluent Radiation Monitoring System I&C Configuration 2 1.7-2 Revision 4.1

re Title re 5.1-2 Reactor Coolant System Simplified Diagram re 5.13 Reactor Coolant System Schematic Flow Diagram re 5.4-9 Steam Generator Simplified Diagram re 5.4-10 Decay Heat Removal System Simplified Diagram re 6.24 Containment System Piping and Instrumentation Diagram re 6.27 Containment Isolation Valve Actuator Hydraulic Schematic re 6.28 Containment Isolation Valve Hydraulic Skid Schematic re 6.3-1 Emergency Core Cooling System re 6.33 Emergency Core Cooling System Valve and Actuator Schematic re 6.4-1 Control Room Habitability System Diagram re 9.1.3-1 Spent Fuel Pool Cooling System Diagram res 9.1.3-2a and 9.1.3-2b Reactor Pool Cooling System Diagram re 9.1.3-3 Pool Cleanup System Diagram re 9.1.3-4 Pool Surge Control System Diagram re 9.2.2-1 Reactor Component Cooling Water System Diagram re 9.2.3-1 Demineralized Water System Diagram re 9.2.5-2 Ultimate Heat Sink Qualified Makeup Line and Instrumentation Diagram re 9.2.61 Condensate Storage Facility re 9.2.7-1 Site Cooling Water System Diagram re 9.2.8-1 Chilled Water System Diagram re 9.2.9-1 Utility Water System Diagram re 9.3.1-1 Instrument Air and Service Air System Diagram re 9.3.1-2 Nitrogen Distribution System Diagram re 9.3.2-1 Containment Sampling System Diagram re 9.3.3-1 Radioactive Waste Drain System Diagram re 9.3.3-2 Balance-of-Plant Drain System Diagram re 9.3.4-1 Chemical and Volume Control System Diagram re 9.3.4-2 Boron Addition System Diagram re 9.3.6-1 Containment Evacuation System Diagram re 9.3.6-2 Containment Flooding and Drain System Diagram re 9.4.1-1 Control Room Ventilation System Diagram re 9.4.2-1 Reactor Building HVAC System Diagram re 9.4.3-1 Radioactive Waste Building HVAC System Diagram re 9.4.4-1 Turbine Building HVAC System Diagram re 9.5.1-1 Fire Protection System Water Supplies and Fire Pumps re 9.5.1-2 Fire Protection System Yard Fire Main Loop re 10.1-1 Power Conversion System Block Flow Diagram re 10.1-2 Flow Diagram and Heat Balance Diagram at Rated Power for Steam and Power Conversion System Cycle re 10.2-1 Turbine Generator System Piping and Instrumentation Diagram re 10.31 Main Steam System Piping and Instrumentation Diagram re 10.4-1 Main Condenser Piping and Instrumentation Diagram re 10.4-2 Condenser Air Removal System Piping and Instrumentation Diagram re 10.4-3 Circulating Water System Piping and Instrumentation Diagram (Typical of 2) res 10.4-4a and 10.4-4b Auxiliary Boiler System Piping and Instrumentation Diagram res 11.2-1a through 11.2-1j Liquid Radioactive Waste System Diagram res 11.3-1a and 11.3-1b Gaseous Radioactive Waste System Diagram re 11.4-1 Block Diagram of the Solid Radioactive Waste System re 11.4-2a Process Flow Diagram for Wet Solid Waste re 11.4-2b Solid Radioactive Waste System Diagram 2 1.7-3 Revision 4.1

re Title re 11.51 Radioactive Effluent Flow Paths with Process and Effluent Radiation Monitors re 11.53 OffLine Radiation Monitor re 11.54 AdjacenttoLine Radiation Monitor re 11.55 InLine Radiation Monitor re 11.56 RBVS Plant Exhaust Stack Effluent Radiation Monitor 2 1.7-4 Revision 4.1

NuScale Final Safety Analysis Report Drawings and Other Detailed Information Figure 1.7-1a: Electrical Symbols SINGLE-LINE DIAGRAM LEGEND

1. POWER TRANSFORMERS 6. CONNECTIONS 9. CONDUCTORS/CABLES 12. SWITCHING 16. LINE TYPES 3 WINDING POWER TRANSFORMER TWO POSITION TRANSFER SWITCH SEE 9. CONDUCTORS/CABLES BRANCH LINE H - HIGH SIDE WINDING CONNECTION STAB CABLE BUS ATS = AUTO TRANSFER SWITCH X,Y - LOW SIDE WINDINGS ATS STATIC = STATIC SWITCH OR OTHER DESCRIPTION SEE 9. CONDUCTORS/CABLES POWER BUS MVA/kVA RATING COOLING METHOD(S) BRANCH LINE IMPEDIENCE RATING COMMON POINT OF NEUTRAL VOLTAGE LINE POTENTIAL 4.16 KV- 480V 4.16KV, 2000A, 50KA POWER BUS 1500 KVA 2 WINDING POWER TRANSFORMER CURRENT LINE RATED VOLTAGE ON/AN PRIMARY-SECONDARY VOLTAGE LEVEL Z= 5.75% CONTINUOUS OPERATING CURRENT MVA/KVA RATING(S) SPARE CONNECTION OR SPECIAL TERMINAL SHORT CIRCUIT SYMMETRICAL INTERRUPT RATING FUSED SAFETY SWITCH COOLING METHOD(S) 40A RELAY ACTUATION, ALARM OR INTERLOCK EX: BATTERY AND/OR CHARGER RIGHT - AMPERE RATING IMPEDANCE RATING TEST TERMINAL CABLE NAME 0-120V 10KVA VOLTAGE REGULATING TRANSFORMER STARTER OR CONTACTOR INSIDE AN ENCLOSURE 1 PRIMARY-SECONDARY VOLTAGE LEVEL 3 LC - LINE CONTACTOR 2000 A CABLE ID TAG SAFETY DISCONNECT SWITCH MVA/KVA RATING(S) LC M -MOTOR TOP RIGHT - AMPERE RATING

                                                               # OF PHASES                                                                      TOP LEFT - SIZE

2. INSTRUMENT TRANSFORMERS NORMALLY OPEN CONTACT 17. TERMINALS CURRENT TRANSFORMER ROTARY SWITCH 200-5A 1 UPPER RIGHT - QTY UPPER LEFT - TRANSFORMER RATIO

10. PROTECTION DEVICES 4 C400 TERMINAL IN STARTER COMPARTMENT LOWER LEFT - ACCURACY CLASS NORMALLY CLOSED CONTACT NUMBER DENOTES TERMINAL NUMBER HIGH VOLTAGE (> 100kV) CIRCUIT BREAKER MULTIPLE POSITION SELECTOR SWITCH N.O. 3000 A RIGHT - AMPERE RATING 400 -120V 3 POTENTIAL TRANSFORMER LEFT - N.O. / BLANK = N.C. 6 TERMINAL IN FIELD TERMINAL BOX UPPER RIGHT - QTY REMOVABLE LINK UPPER LEFT - TRANSFORMER RATIO NUMBER DENOTES TERMINAL NUMBER 2 POSITION SELECTOR SWITCH SHUNT 8 POTENTIAL TRANSFORMER TERMINAL IN LOCAL CONTROL PANEL 400 - 120 V 1 DOUBLE SECONDARY MEDIUM VOLTAGE (1kV - 100kV) CIRCUIT BREAKER 400 - 120 V NUMBER DENOTES TERMINAL NUMBER UPPER RIGHT - QTY RIGHT - AMPERE RATING UPPER LEFT - TRANSFORMER RATIO CONNECTION LEFT - N.O. / BLANK = N.C. 2000A DISCONNECT SWITCH WITH INTERLOCKED 1200 A 52 BOTTOM - BREAKER DESIGN M (TO BUS, CABLE TERMINAL OR OTHER). N.O. 6A TOP - DEVICE TYPE GROUNDING SWITCH 25-1 TERMINAL IN DCS/PLC PANEL

3. TRANSFORMER WINDING DESIGNATIONS LEFT - AMPERE RATING NUMBER DENOTES TERMINAL NUMBER DELTA CONNECTED TRANSFORMER WINDINGS LINES CROSSING (NO CONNECTION)

13. INSTRUMENTATION 7 TERMINAL IN LOW VOLTAGE LOW VOLTAGE (<1kV) AC OR DC CIRCUIT BREAKER 200 A MOTOR CONTROL CENTER (MCC)

WYE CONNECTED TRANSFORMER WINDINGS LINE CONTINUATION N.O. N.O. 200 A RIGHT - AMPERE RATING XX LEFT - N.O. / BLANK = N.C. METER OR INSTRUMENT (INSIDE) NUMBER DENOTES TERMINAL NUMBER (XX - DENOTES ELECTRICAL CONNECTION A A - AMMETER BETWEEN SYMBOLS BEARING SAME F - FREQUENCY METER A CONNECTION ID) SYN - SYNCHROSCOPE C B PHASOR INDICATION V - VOLTMETER OR VAR - VAR METER 5 3 FUSE TERMINAL IN LOW VOLTAGE SWITCH GEAR 200A W - WATT METER

4. GROUNDING STYLE DESIGNATIONS RIGHT - AMPERE RATING NUMBER DENOTES TERMINAL NUMBER TI - TEMPERATURE INDICATOR LEFT - QUANTITY DMM - DIGITAL MULTIMETER CONNECTING DRAWING IDENTIFIER IDENTIFIER CONNECTING DRAWING NP12-YY-DYY-E-SL-YYY-SYY XXXXX XXXXX NP12-YY-DYY-E-SL-YYY-SYY

                                        .25KVA            NEUTRAL GROUNDING TRANSFORMER                                                                                                                                                                                                                                              MONITORING DEVICE TO/FROM ZZZ                                          TO/FROM ZZZ                                                                                                                                   BM 19000-120/240V                                                                                                                                                                                                  THERMAL OVERLOAD ELEMENT                                                            BM - BATTERY MONITOR CONNECTING EQUIPMENT                               CONNECTING EQUIPMENT                                                                                                                                                                                                                2                TERMINAL IN MEDIUM VOLTAGE SWITCH GEAR NUMBER DENOTES TERMINAL NUMBER 51               RELAY/DEVICE

7. LOADS 3 LIGHTNING ARRESTER G SEE NP-ES-0303-2923 WYE CONNECTED TRANSFORMER 18. MISCELLANEOUS KV RIGHT - VOLTAGE RATING FOR RELAY/DEVICE TYPES.

WINDINGS WITH GROUND TOP LEFT - QUANTITY 20 KW ELECTRICAL LOAD WITH KW RATING 14. CONTROL DEVICES SURGE CAPACITOR DELTA WYE TRANSFORMER WITH LOAD NAME RESISTANCE GROUND 11. DC POWER VFD VARIABLE FREQUENCY DRIVE ELECTRICAL HEATER 2 HP MOTOR WITH HORSEPOWER RATING TOP - KW RATING INVERTER ABBREVIATIONS INPUT: 125 VDC MOTOR NAME DC INPUT VOLTAGE OUTPUT: 120/208 VAC, 3 LOCAL LOCAL CONTROLLER GROUND SYMBOL AC OUTPUT VOLTAGE, # N.O. - NORMALLY OPEN

8. GENERATORS S CLASS BREAK N.C. - NORMALLY CLOSED

15. INTERLOCKS CLASS BREAK S - SAFETY QTY - QUANTITY NS - NONSAFETY

5. TRANSFORMER LABELS SYNCHRONOUS GENERATOR NS SWGR - SWITCH GEAR INPUT: 480 VAC, 3 CHARGER 60 HZ OPERATING FREQUENCY OUTPUT: 125 VDC AC INPUT VOLTAGE, #

1800 RPM OPERATING RPM K - KEY DC OUTPUT VOLTAGE K MPT MAIN POWER TRANSFORMER 90% PF  % POWER FACTOR M M - MECHANICAL UAT UNIT AUXILIARY TRANSFORMER MULTICELL BATTERY 125 VDC 150 AH VOLTAGE RATING E E - ELECTRICAL AMP HOUR RATING 60 CELLS SST STATION SERVICE TRANSFORMER NUMBER OF CELLS Tier 2 1.7-5 Revision 4.1

NuScale Final Safety Analysis Report Drawings and Other Detailed Information Figure 1.7-1b: Electrical Symbols OTHER ELECTRICAL DIAGRAM LEGEND

19. SWITCHES 22. ELECTRONICS 23. )6$5&+$37(563(&,),&6<0%2/6$1'127(6 20 PRESSURE SWITCH (NC) RESISTOR (FIXED)

PRESSURE SWITCH (NO) RESISTOR VARIABLE TEMPERATURE SWITCH (NO) DIODE LEVEL SWITCH (NC) GTO SCR LEVEL SWITCH (NO) THYRISTOR FLOW SWITCH (NC) ZENER DIODE FLOW SWITCH (NO) ARC SUPPRESSOR LIMIT SWITCH (NOHC) TERMINAL BOARD NORMALLY OPEN HELD CLOSED LIMIT SWITCH (NCHO) SOLENOID NORMALLY CLOSED HELD OPEN TIMER CONTACT (NCTC) NORMALLY CLOSED TIME CLOSED COIL RELAY MOTOR GENERATOR FIELD SPEED SWITCH CPT CONTROL POWER TRANSFORMER PRIMARY - SECONDARY VOLTAGE LEVEL 480V-120V

20. PUSH BUTTONS RTD RESISTANCE TEMPERATURE DETECTOR PUSH BUTTON, NORMALLY CLOSED (NC) 3 480 VAC WELDING RECEPTACLE 100 A PUSH BUTTON, NORMALLY OPENED (NO)

PUSH BUTTON, 2 POSITION PUSH-TO-TEST INDICATING LIGHT R

21. INDICATING/ALARM ABBREVIATIONS INDICATING LIGHT A A = AMBER, B = BLUE C = CLEAR, G = GREEN DCS - DISTRIBUTED CONTROL SYSTEM R = RED , W = WHITE GTO - GATE TURN OFF NO - NORMALLY OPEN NC - NORMALLY CLOSED HORN PLC - PROGRAMMABLE LOGIC CONTROLLER SCR - SILICON CONTROLLED RECTIFIER Tier 2 1.7-6 Revision 4.1

NuScale Final Safety Analysis Report Drawings and Other Detailed Information Figure 1.7-2: Instrumentation and Controls Symbol Legend THE OR GATE OUTPUT IS TRUE IF ANY THE OUTPUT VALUE IS A NONLINEAR OR INPUT IS TRUE. UNSPECIFIED FUNCTION OF THE INPUT. THE FUNCTION IS DEFINED IN A NOTE OR OTHER TEXT. THE AND GATE OUTPUT IS TRUE ONLY THE OUTPUT VALUE IS THE ALGEBRAIC SUM IF ALL INPUTS ARE TRUE. OF THE INPUTS. THE AND GATE OUTPUT IS TRUE ONLY IF 2 OUT OF 4 INPUTS, OR MORE, ARE TRUE. THE NOT GATE OUTPUT IS FALSE IF THE INPUT IS TRUE. THE OUTPUT IS TRUE IF THE INPUT IS FALSE. A FUNCTION WHICH PRODUCES AN OUTPUT FOLLOWING A PRESET TIME DELAY AFTER RECEIVING AN INPUT. IF THE INPUT CHANGES FROM TRUE TO FALSE BEFORE THE PRESET TIME DELAY ELAPSES, THEN THE OUTPUT REMAINS FALSE BISTABLE OUTPUT IS A LOGIC 1 WHEN THE MEASURED VARIABLE IS GREATER THAN THE SETPOINT VALUE. BISTABLE OUTPUT IS A LOGIC 1 WHEN THE MEASURED VARIABLE IS LESS THAN THE SETPOINT VALUE. ONE OF FOUR REDUNDANT SEPARATION GROUPS (SGS) SGS ARE IDENTIFIED AS A, B, C, OR D. MOMENTARY HAND SWITCH LOCATED IN THE MAIN CONTROL ROOM . MAINTAINED-POSITION HAND SWITCH LOCATED IN THE MAIN CONTROL ROOM . MAIN CONTROL ROOM INDICATION OF SYSTEM TRIP/ACTUATION STATUS. MAIN CONTROL ROOM INDICATION OF OPERATIONAL BYPASS PERMISSIVE STATUS. MAIN CONTROL ROOM ANNUNCIATOR FOR SYSTEM TRIP/ACTUATION STATUS. MAIN CONTROL ROOM FIRST OUT ANNUNCIATOR FOR SYSTEM TRIP/ ACTUATION STATUS. LOCALLY MOUNTED SWITCH FOR MAINTENANCE AND TESTING. MCS MOMENTARY MANUAL CONTROL INTERFACE LOCATED IN THE MAIN CONTROL ROOM FOR NONSAFETY CONTROL OF MPS ACTUATED EQUIPMENT. Tier 2 1.7-7 Revision 4.1

NuScale Final Safety Analysis Report Drawings and Other Detailed Information Figure 1.7-3a: Piping and Instrumentation Diagram Legends VALVES FIRE & SAFETY SPECIALTY ITEMS FITTINGS/MISC DRAINS NOTES: ALL VENTS & DRAINS ON THE P&ID BALL VALVE NEEDLE VALVE FIRE HYDRANT FLAME ARRESTOR CONCENTRIC REDUCER REPRESENT A 0.75" NORMALLY CLOSED 1. ARROW INDICATES FAILURE OR UNACTUATED LOOP SEAL VALVE. EXCEPTIONS MUST BE NOTED ON FLOW PATH. P&ID. THREE-WAY BALL VALVE FIRE HYDRANT W/HOSE HOUSE HAMMER ARRESTOR OIL SEPARATOR ECCENTRIC REDUCER (NOTE 1) KNIFE GATE VALVE CLOSED VENT FOUR-WAY, FOUR-PORTED FREEZE PROOF YARD HYDRANT OR DRAIN EDUCTOR CAP (BUTT WELD) BALL VALVE EXPANSION JOINT XXXX = SYSTEM SLIDE VALVE TO XXXX (NOTE 1) ABBREVIATION FREEZE PROOF HOSE VALVE EJECTOR HINGED EXPANSION JOINT SCREWED END STRAIGHT GLOBE VALVE OPEN VENT THREE-WAY SLIDE VALVE KEY OPERATED VALVE SWIVEL JOINT INLINE SIGHT GLASS HOSE CONNECTION OR DRAIN ANGLED GLOBE VALVE ALM TO XXXX XXXX = SYSTEM T POST INDICATOR VALVE SINGLE BASKET STRAINER BREATHER VENT ABBREVIATION W/ TAMPER SWITCH CAPPED HOSE CONNECTION THREE-WAY GLOBE VALVE FLOAT VALVE (NOTE 1) AR DUPLEX BASKET STRAINER SPRAY NOZZLE FLANGE Y BLOWDOWN VALVE AIR RELEASE VALVE FLOOR DRAIN SAFETY ANGLE VALVE / GENERIC SIMPLEX BASKET STRAINER T STEAM TRAP BLIND FLANGE TWO-WAY ANGLE VALVE ANGLE BLOWDOWN VALVE DRY PIPE VALVE DRAIN FUNNEL SUMP STRAINER THREE-WAY GATE VALVE / GENERIC IN-LINE-MIXER REDUCING FLANGE THREE-WAY VALVE D DELUGE VALVE (NOTE 1) AUTOMATIC RECIRCULATION VALVE T-STRAINER MIXING TEE CLEAN OUT CLOSED SPECTACLE FLANGE FOAM CHAMBER STARTUP STRAINER IN-LINE SILENCER FOUR-WAY, FOUR-PORTED GATE VALVE / DRAIN GENERIC FOUR-WAY VALVE (NOTE 1) VALVE STEM EXTENDED THROUGH VENT SILENCER OPEN SPECTACLE FLANGE SHIELDED WALL HOSE RACK STATION SCREEN STRAINER GATE VALVE / GENERIC TWO-WAY IN-LINE SAMPLER BACKFLOW PREVENTER S VALVE SINGLE BLIND GATE VALVE, FLANGED HOSE REEL CONE STRAINER ISOKINETIC SAMPLER FOUR-WAY, FIVE-PORTED VALVE FOOT VALVE RING SPACER (NOTE 1) FIRE MONITOR TEMPORARY STRAINER EXTENDED BODY GATE VALVE PULSATION DAMPENER BUTTERFLY VALVE OPEN & CLOSED SPECTACLE VENT GATE VALVE, PLUGGED FLANGES WITH PIPE FLANGES HOSE VALVE ELEVATED FIRE MONITOR BUTTERFLY VALVE NC SIPHON VACUUM BREAKER ANGLED HOSE VALVE REMOTELY OPERATED FIRE SINGLE BLIND WITH PIPE PLUG VALVE Y-STRAINER MONITOR FLANGES BALL VALVE, FLANGED CIRCUIT SETTER THREE-WAY PLUG VALVE INJECTION ELEMENT (NOTE 1) FOAM MONITOR FILTER AUTOMATIC VENT VALVE BALL VALVE, PLUGGED FOUR-WAY PLUG VALVE FLEXIBLE HOSE (NOTE 1) ELEVATED FOAM MONITOR DISTRIBUTOR BREAK POT THREADED PLUG SKIMMER ECCENTRIC ROTARY DISC VALVE INLET AIR FILTER ORIFICE PLATE DIAPHRAGM VALVE REMOTELY OPERATED FOAM UNION MONITOR RS REMOVABLE SPOOL TEST PORT DIAPHRAGM VALVE DIELECTRIC UNION SAFETY SHOWER MECHANICAL COUPLING FLOW NOZZLE PINCH VALVE DIELECTRIC FLANGE DRIP LEG SAFETY SHOWER W/ EYE WASH WALL PENETRATION BELLOWS SEALED VALVE PITOT TUBE AVERAGING ROOF, FLOOR, OR GROUND CHECK VALVE PENETRATION EYE WASH EXHAUST VENT BALL JOINT CHECK VALVE WITH 3/32 SWING ELBOW ORIFICE IN CLAPPER PRE-ACTION SPRINKLER FREE VENT WITH SCREEN STOP CHECK VALVE SPRAY SPRINKLER RUPTURE DISK FREE VENT WITHOUT SCREEN WAFER CHECK VALVE WET SPRINKLER SAMPLE COOLER TILTING DISC CHECK VALVE ALM S SAMPLER BUTTERFLY VALVE T W/ TAMPER SWITCH SPRAY DESUPERHEATER ANGLE CHECK VALVE ALM GATE VALVE LIFT CHECK VALVE T DESUPERHEATER W/ TAMPER SWITCH EXCESS FLOW CHECK VALVE PACKED BED STEM LEAK-OFF VALVE TRIPLE DUTY VALVE Tier 2 1.7-8 Revision 4.1

NuScale Final Safety Analysis Report Drawings and Other Detailed Information Figure 1.7-3b: Piping and Instrumentation Diagram Legends ACTUATORS SELF-ACTUATING FINAL CONTROL ELEMENTS ADDITIONAL ACTUATOR TYPES PNEUMATIC ACTUATOR PRESSURE SETTING (GENERIC ACTUATOR / SPRING-DIAPHRAGM ACTUATOR) XXX AUTOMATIC FLOW REGULATOR XXX=FCV WITHOUT INDICATOR PSET = 100 psig BALL FLOAT XXX=FICV WITH INTEGRAL INDICATOR PRESSURE/VACUUM RELIEF MANHOLE COVER PNEUMATIC ACTUATOR (SPRING-DIAPHRAGM ACTUATOR W/ POSITIONER) CAPACITANCE SENSOR (A) VARIABLE AREA FLOWMETER W/ INTEGRAL MANUAL PNEUMATIC ACTUATOR ADJUSTING VALVE PRESSURE-REDUCING REGULATOR W/ INTEGRAL OUTLET (PRESSURE-BALANCED DIAPHRAGM ACTUATOR) (INSTRUMENT TAG BUBBLE REQUIRED WITH 'B') PG PRESSURE RELIEF AND PRESSURE GAUGE. DISPLACEMENT FLOAT LINEAR PISTON ACTUATOR (SINGLE-ACTING, SPRING-OPPOSED OR DOUBLE-ACTING) PRESSURE SETTING PADDLE WHEEL CONSTANT FLOW REGULATOR HYDRAULIC CONTROL LINEAR PISTON ACTUATOR WITH POSITIONER FICV PSET = 100 psig ANGLE PRESSURE RELIEF VALVE / GENERIC PRESSURE SAFETY HYDRAULIC ACTUATOR, VALVE REMOTELY OPERATED W/ LOCAL H HYDRAULIC ACTUATOR N2 TANK FOR SAFE POSITIONING. ROTARY PISTON ACTUATOR (SINGLE-ACTING, SPRING-OPPOSED OR DOUBLE-ACTING) FG FLOW SIGHT GLASS (TYPE SHALL BE NOTED IF MORE THAN ONE TYPE IS USED) SEQUENCING VALVE ROTARY PISTON ACTUATOR WITH POSITIONER ANGLE VACUUM RELIEF VALVE / GENERIC VACUUM SAFETY FO GENERIC FLOW RESTRICTION / SINGLE STAGE ORIFICE VALVE PLATE REDUCING VALVE BELLOWS SPRING OPPOSED ACTUATOR (NOTE REQUIRED FOR MULTI-STAGE OR CAPILLARY TUBE TYPES) M ROTARY MOTOR-OPERATED ACTUATOR PRESSURE COMPENSATED FO STRAIGHT-THRU PRESSURE RELIEF VALVE FLOW CONTROL RESTRICTION ORIFICE DRILLED IN VALVE PLUG (TAG NUMBER SHALL BE OMITTED IS VALVE IS SOLENOID ACTUATOR (OPEN-CLOSE OR MODULATING) OTHERWISE IDENTIFIED) 2 POS, 4 WAY DIRECTIONAL S CONTROL VALVE, SOLENOID OPERATED WITH SPRING RETURN. PRESSURE-VACUUM RELIEF VALVE ACTUATOR WITH SIDE-MOUNTED HANDWHEEL &21752/9$/9()$,/85(326,7,216 TANK 3 POS, 4 WAY, CLOSED CENTER LEVEL REGULATOR W/ BALL FLOAT AND DIRECTIONAL CONTROL VALVE, MECHANICAL LINKAGE )$,/7223(1326,7,21 SOLENOID OPERATED. ACTUATOR WITH TOP-MOUNTED HANDWHEEL PRESSURE SAFETY ELEMENT / PRESSURE RUPTURE DISK 3 POS, 4 WAY, DRAIN CENTER DIRECTIONAL CONTROL VALVE, MANUAL ACTUATOR )$,/72&/26('326,7,21 SOLENOID OPERATED .. BACKPRESSURE REGULATOR, INTERNAL PRESSURE TAP VACUUM SAFETY ELEMENT / VACUUM RUPTURE DISK 3 POS, 4 WAY, FLOAT CENTER ACTUATOR WITH MANUAL ACTUATED PARTIAL STROKE DIRECTIONAL CONTROL VALVE, TEST DEVICE SOLENOID OPERATED..

                                                                                                                                                                                                                                                                             )$,//2&.('$6,6 ACTUATOR WITH REMOTE ACTUATED PARTIAL STROKE                            BACKPRESSURE REGULATOR, EXTERNAL PRESSURE TAP                                               TEMPERATURE REGULATOR W/ FILLED THERMAL SYSTEM                                                                           TO RESERVOIR S     TEST DEVICE PRESSURE INTENSIFIER ON-OFF SOLENOID ACTUATOR, NON-LATCHING / AUTOMATIC RESET                                                                                                                                                                                      )$,/$6,6'5,)723(1 S

THREE-WAY TEMPERATURE REGULATOR W/ FILLED THERMAL SYSTEM CYLINDER S ON-OFF SOLENOID ACTUATOR, LATCHING / ON-OFF SOLENOID PRESSURE REDUCING REGULATOR, INTERNAL PRESSURE TAP R ACTUATOR, MANUAL OR REMOTE RESET

                                                                                                                                                                                                                                                                             )$,/$6,6'5,)7&/26('

TANDEM CYLINDER S ON-OFF SOLENOID ACTUATOR, LATCHING / ON-OFF SOLENOID R R ACTUATOR, MANUAL AND REMOTE RESET HYDRAULIC MOTOR SPRING OR WEIGHT ACTUATED RELIEF OR SAFETY VALVE ANGLED TEMPERATURE REGULATOR W/ FILLED THERMAL SYSTEM ACTUATOR PRESSURE REDUCING REGULATOR, EXTERNAL PRESSURE TAP ADDITIONAL VALVE STATUS INFORMATION P PILOT ACTUATED RELIEF OR SAFETY VALVE ACTUATOR W/ PRESSURE SENSING LINE. OPEN DURING NORMAL OPERATION (PILOT PRESSURE SENSING LINE DELETED IF DIFFERENTIAL PRESSURE REGULATOR, EXTERNAL PRESSURE (ALL VALVES EXCEPT BUTTERFLY VALVES) SENSING IS INTERNAL) TAPS TSE THERMAL SAFETY ELEMENT, FUSIBLE PLUG OR DISK CLOSED DURING NORMAL OPERATION (ALL VALVES EXCEPT BUTTERFLY VALVES) E ELECTRIC TO PNEUMATIC CONTROL P DIFFERENTIAL PRESSURE REGULATOR, INTERNAL PRESSURE SIGNAL CONVERTER TAPS STEAM TRAP / GENERIC MOISTURE TRAP T XX ACTION E ELECTRIC TO HYDRAULIC CONTROL (NOTE REQUIRED FOR OTHER TRAP TYPES) H NO - NORMALLY OPEN SIGNAL CONVERTER NC - NORMALLY CLOSED DIFFERENTIAL PRESSURE REGULATOR W/ INTERNAL & EXTERNAL FO - FAIL OPEN E VOLTAGE TO CURRENT CONTROL PRESSURE TAPS FC - FAIL CLOSED I TANK SIGNAL CONVERTER MOISTURE TRAP WITH EQUALIZATION LINE FL - FAIL LAST T LO - LOCKED OPEN PRESSURE REDUCING REGULATOR, INTERNAL PRESSURE TAP LC - LOCKED CLOSED E ELECTROHYDRAULIC LINEAR OR WITH GLOBE VALVE H ALL INLINE VALVES ON P&ID ARE ROTARY ACTUATOR NORMALLY OPEN. EXCEPTIONS MUST BE NOTED ON P&ID. Tier 2 1.7-9 Revision 4.1

NuScale Final Safety Analysis Report Drawings and Other Detailed Information Figure 1.7-3c: Piping and Instrumentation Diagram Legends TANKS, VESSELS, HOPPERS NOZZLE TYPES N1 MANWAY (MW) HVAC COILS N2 BUTTWELD NOZZLE HORIZONTAL VESSEL DEAERATOR HIGH EFFICIENCY C H N3 FLANGED NOZZLE HIGH AHU FILTER HUMIDIFIER COOLING COIL, HEATING COIL, CHILLED WATER HOT WATER CW HW N4 FLANGED NOZZLE W/BLIND MEDIUM EFFICIENCY MED DEHUMIDIFIER H H AHU FILTER HEATING COIL, HEAT HEATING COIL, RECOVERY ELECTRIC N5 FLANGED NOZZLE - ANGLED HR EH LOW EFFICIENCY HOPPER LOW CENTRIFUGAL FAN AHU FILTER D H DIRECT EXPANSION HEATING COIL, CHEMICAL TOTE COIL STEAM VERTICAL VESSEL VESSEL JACKETED VESSEL VERTICAL DRUM X ST CHAR CHARCOAL FILTER AHU FAN REHEAT COIL HIGH-EFFICIENCY HEPA PARTICULATE AIR FILTER PARALLEL BLADE VAV VARIABLE AIR VOLUME BOX - DAMPER W/ REHEAT COIL T TEST SECTION E FOR FILTERS W/ S T MANIFOLD OPPOSED BLADE DAMPER M TANK WITH CONE ROOF TANK WITH DOME ROOF TANK WITH INTERNAL FLOATING ROOF TANK WITH DIAPHRAGM CONE BOTTOM TANK I MIXING BOX X HEAT EXCHANGERS & HEAT TRANSFER EQUIPMENT S SECURITY BARRIER B COMPRESSOR F FIRE DAMPER D SLIDE GATE DAMPER EVAPORATOR S SMOKE DAMPER WALL WALL D SHELL & TUBE HEAT EXCHANGER SHELL & TUBE HEAT DOUBLE PIPE HEAT EXCHANGER TUBE HEAT EXCHANGER HEAT EXCHANGER EVAPORATOR MOUNTED MOUNTED COMPRESSOR GENERIC BLADE UNIT HEATER AC UNIT EXCHANGER (END VIEW) CHILLER DAMPER DHR HEAT EXCHANGER BLAST DAMPER AIR FLOW UNIT HEATER CONDENSER EVAPORATOR TORNADO DAMPER EXHAUST FAN EXHAUST FAN AIR FLOW CEILING MOUNTED AXIAL UPBLAST LOUVER HELICAL COIL HEATER EXTERNAL AIR COOLER COOLING TOWER STEAM GENERATOR ELECTRICAL JACKET PLATE & FRAME COOLING TOWER CELL CHILLER HEAT EXCHANGER VARIABLE AIR VAV PROPELLER FAN VOLUME BOX PUMPS COMPRESSORS & BLOWERS FILTERS MISCELLANEOUS DRIVERS CONDENSER EVAPORATOR ADJUSTABLE SPEED MANUAL ASD DRIVE (MECHANICAL) OR VERTICAL POSITIVE PROGRESSIVE VERTICAL SUBMERSIBLE AIR COOLED LIQUID TYPE BAG TYPE RAKE STEAM TURBINE PUMP DISPLACEMENT CAVITY PUMP INLINE PUMP SUMP PUMP CONDENSING UNIT FILTER FILTER FILTER PUMP M ELECTRIC MOTOR VARIABLE FREQUENCY VFD DRIVE (ELECTRICAL) PERMANENT MAGNET SKIMMER ELECTRIC MOTOR VARIABLE SPEED VERTICAL CAN CENTRIFUGAL METERING TURBINE SCREW PUMP COUPLING POSITIVE RECIPROCATING CENTRIFUGAL AGITATOR PUMP SUMP PUMP PUMP PUMP DISPLACEMENT COMPRESSOR COMPRESSOR STEAM TURBINE PNEUMATIC BLOWER D DIESEL ENGINE HYDRAULIC HORIZONTAL LIQUID RING DIAPHRAGM PRESSURE SCREW OIL CENTRIFUGAL VACUUM PUMP COMPRESSOR COMPRESSOR SEPARATOR PUMP PUMP VARIABLE VOLUME PUMP Tier 2 1.7-10 Revision 4.1

NuScale Final Safety Analysis Report Drawings and Other Detailed Information Figure 1.7-3d: Piping and Instrumentation Diagram Legends INSTRUMENTATION DEVICE AND FUNCTION SYMBOLS 127( NOTES: SHARED DISPLAY, SHARED CONTROL PRIMARY CHOICE OR BASIC ALTERNATE CHOICE OR SAFETY COMPUTER SYSTEMS DISCRETE LOCATION AND ACCESSIBILITY  INSTRUMENT FUNCTION IDENTIFIER. USED PROCESS CONTROL SYSTEM INSTRUMENTED SYSTEM AND SOFTWARE ONLY WHEN THE COMPONENT TYPE REQUIRES FURTHER CLARIFICATION. SEE x LOCATED IN FIELD FUNCTION IDENTIFIERS TABLE ON ),*85( x NOT PANEL, CABINET OR CONSOLE MOUNTED H. x VISIBLE AT FIELD LOCATION x NORMALLY OPERATOR ACCESSIBLE

                                                                                                                                                                                                                                                                                                                                                               )25,167580(17$7,217<3($1'

x LOCATED IN OR ON FRONT OF CONTROL OR MAIN PANEL OR CONSOLE INSTRUMENT TAGGING

                                                                                                                                                                                                                                                                                                                                                                 )81&7,216((,167580(17$7,21

x VISIBLE ON FRONT OF PANEL OR ON VIDEO DISPLAY ,'(17,),&$7,21/(77(567$%/(217+,6 x NORMALLY OPERATOR ACCESSIBLE AT PANEL FRONT OR CONSOLE COMPONENT TYPE 6+((7$1')81&7,21,'(17,),(567$%/( 21),*85(H UNIQUE IDENTIFIER x LOCATED IN REAR OF CENTRAL OR MAIN PANEL XXX x LOCATED IN CABINET BEHIND PANEL  TEMPERATURE ELEMENTS ARE XXXX x NOT VISIBLE ON FRONT OF PANEL OR ON VIDEO DISPLAY XXX THERMOCOUPLES UNLESS NOTED x NOT NORMALLY OPERATOR ACCESSIBLE AT PANEL OR CONSOLE INSTRUMENT FUNCTION OTHERWISE. IDENTIFIER (NOTE 1) x LOCATED IN OR ON FRONT OF SECONDARY OR LOCAL PANEL OR CONSOLE  FOR GUIDANCE ON THE USE OF THE x VISIBLE ON FRONT OF PANEL OR ON VIDEO DISPLAY INSTRUMENTATION IDENTIFICATION x NORMALLY OPERATOR ACCESSIBLE AT PANEL FRONT OR CONSOLE LETTERS TABLE, REFER TO ANSI/ISA 5.1-2009. x LOCATED IN REAR OF SECONDARY OR LOCAL PANEL x LOCATED IN FIELD CABINET x NOT VISIBLE ON FRONT OF PANEL OR ON VIDEO DISPLAY x NOT NORMALLY OPERATOR ACCESSIBLE AT PANEL OR CONSOLE TYPICAL INSTRUMENT COMPONENT COMBINATIONS INSTRUMENTATION IDENTIFICATION LETTERS (NOTE ) READOUT FIRST LETTERS SUCCEEDING LETTERS SOLENOIDS, CONTROLLERS DEVICES SWITCHES AND ALARM DEVICES* TRANSMITTER RELAYS, VIEWING COLUMN 1 COLUMN 2 COLUMN 3 COLUMN 4 COLUMN 5 FIRST CONTROL COMPUTING PRIMARY TEST WELL OR DEVICE, SAFETY FINAL MEASURED/INITIATING VARIABLE VARIABLE READOUT/PASSIVE OUTPUT/ACTIVE FUNCTION INDICATING OR MEASURABLE VARIABLE INDICATING BLIND INDICATING HIGH** LOW** COMB** INDICATING BLIND LETTERS VALVES DEVICES ELEMENT POINT PROBE GLASS DEVICE ELEMENT MODIFIER FUNCTION FUNCTION MODIFIER A ANALYSIS AIC AC AI ASH ASL ASHL AIT AT AY AE AP AW AV A ANALYSIS ALARM B BURNER, COMBUSTION USER'S CHOICE USER'S CHOICE USER'S CHOICE B BURNER/COMBUSTION BIC BC BI BSH BSL BSHL BIT BT BY BE BW BG BZ C USER'S CHOICE CONTROL CLOSE C USER'S CHOICE D USER'S CHOICE DIFFERENCE, DIFFERENTIAL DEVIATION D USER'S CHOICE E VOLTAGE SENSOR, PRIMARY ELEMENT E VOLTAGE EIC EC EI ESH ESL ESHL EIT ET EY EE EZ F FLOW, FLOW RATE RATIO F FLOW RATE FIC FC FCV FI FSH FSL FSJ FIT FT FY FE FP FG FV G USER'S CHOICE GLASS, GAUGE, VIEWING DEVICE FQ FLOW QUANTITY FQIC FQI FQSH FQSL FQV H HAND HIGH FF FLOW RATIO FFIC FFC FFI FFSH FFSL I CURRENT INDICATE J POWER SCAN G USER'S CHOICE K TIME, SCHEDULE TIME RATE OF CHANGE CONTROL STATION H HAND HIC HC HS HV L LEVEL LIGHT LOW I CURRENT IIC II ISH ISL ISHL IIT IT IY IE IZ M USER'S CHOICE MIDDLE, INTERMEDIATE J POWER JIC JI JSH JSL JSHL JIT JT JY JE JZ N USER'S CHOICE USER'S CHOICE USER'S CHOICE USER'S CHOICE K TIME KIC KC KCV KI KSH KSL KSHL KIT KT KY KE KZ O USER'S CHOICE ORIFICE, RESTRICTION OPEN L LEVEL LIC LC LCV LI LSH LSL LSHL LIT LT LY LE LW LG LV P PRESSURE POINT (TEST CONNECTION) M USER'S CHOICE Q QUANTITY INTEGRATE, TOTALIZE INTEGRATE, TOTALIZE R RADIATION RECORD RUN N USER'S CHOICE S SPEED, FREQUENCY SAFETY SWITCH STOP O USER'S CHOICE T TEMPERATURE TRANSMIT P PRESSURE/VACUUM PIC PC PCV PI PSH PSL PSHL PIT PT PY PE PP PSV PV U MULTIVARIABLE MULTIFUNCTION MULTIFUNCTION PD PRESSURE, DIFFERENTIAL PDIC PDC PDCV PDI PDSH PDSL PDIT PDT PDY PDE PDP PSE PDV V VIBRATION, MECHANICAL ANALYSIS VALVE, DAMPER, LOUVER Q QUANTITY QIC QI QSH QSL QSHL QIT QT QY QE QZ W WEIGHT, FORCE WELL, PROBE R RADIATION RIC RC RI RSH RSL RSHL RIT RT RY RE RW RZ X UNCLASSIFIED X-AXIS ACCESSORY DEVICES, UNCLASSIFIED UNCLASSIFIED UNCLASSIFIED S SIC SC SCV SI SSH SSL SSHL SIT ST SY SE SV Y EVENT, STATE, PRESENCE Y-AXIS AUXILIARY DEVICES SPEED/FREQUENCY Z POSITION, DIMENSION Z-AXIS, SAFETY DRIVER, ACTUATOR, T TEMPERATURE (NOTE ) TIC TC TCV TI TSH TSL TSHL TIT TT TY TE TP TW TSE TV INSTRUMENTED SYSTEM UNCLASSIFIED FINAL TD TEMPERATURE, DIFFERENTIAL TDIC TDC TDCV TDI TDSH TDSL TDIT TDT TDY TDE TDP TDW TCV CONTROL ELEMENT U MULTI VARIABLE UI UV V VIBRATION/MACHINERY ANALYSIS VI VSH VSL VSHL VIT VT VY VE W WEIGHT/FORCE WIC WC WCV WI WSH WSL WSHL WIT WT WY WE WZ WD WEIGHT/FORCE,DIFFERENTIAL WDIC WDC WDCV WDI WDSH WDSL WDIT WDT WDY WDE WDZ X UNCLASSIFIED XZ Y EVENT/STATE/PRESENCE YIC YC YI YSH YSL YT YY YE YZ Z POSITION ZIC ZC ZCV ZI ZSH ZSL ZSHL ZIT ZT ZY ZE ZV ZD GAUGING/DEVIATION ZDIC ZDC ZDCV ZDI ZDSH ZDSL ZDIT ZDT ZDY ZDE ZDV NOTE: THIS TABLE IS NOT ALL-INCLUSIVE OTHER POSSIBLE COMBINATIONS:

                  *A, ALARM, THE ANNUNCIATION DEVICE, MAY BE USED IN THE SAME FASHION AS S, SWITCH, THE ACTUATION DEVICE                                  FO      (RESTRICTION ORIFICE)             QQI       (INDICATING COUNTER)
                  ** THE LETTERS "H" AND "L" MAY BE OMITTED IF NOT DEFINED. IF APPROPRIATE, "C" (CLOSED) AND                                              PFR     (PRESSURE RATIO RECORD)           HCV       (HAND CONTROL VALVE)
                      "O" (OPEN) MAY BE USED IN PLACE OF "H" AND "L."                                                                                     KQI     (TIME TOTALIZING INDICATOR)

Tier 2 1.7-11 Revision 4.1

NuScale Final Safety Analysis Report Drawings and Other Detailed Information Figure 1.7-3e: Piping and Instrumentation Diagram Legends FUNCTION IDENTIFIERS INSTRUMENTATION LINE SYMBOLS: INSTRUMENT TO PROCESS AND EQUIPMENT CONNECTIONS ANALYSIS SYMBOL DESCRIPTION CALOR CALORIMETER GC GAS CHROMATOGRAPH MeOH METHYL ALCOHOL ORP OXIDATION REDUCTION TDS TOTAL DISSOLVED SOLIDS 127(6 CO CARBON MONOXIDE H2 HYDROGEN/HYDROGEN ANALYSIS MOIST MOISTURE PHASE PHASE THC TOTAL HYDROCARBON x INSTRUMENT CONNECTION TO PROCESS AND EQUIPMENT CO2 CARBON DIOXIDE HC HYDROCARBON MS MASS SPECTROMETER pH HYDROGEN ION TOC TOTAL ORGANIC CARBON x PROCESS IMPULSE LINES x ANALYZER SAMPLE LINES  63(&,),&$7,216/,67(')259$5,286 COL COLOR H2O WATER NIR NEAR INFRARED REF REFRACTOMETER TURB TURBIDITY (48,30(177<3(6$5(5(&200(1'$7,216 COMB COMBUSTION H2S HYDROGEN SULFIDE N2 NITROGEN RI REFRACTIVE INDEX UV ULTRAVIOLET x HEAT (COOL) TRACED IMPULSE OR SAMPLE LINE FROM PROCESS 21/<7+(5(48,5('63(&,),&$7,216$5( COND ELECTRICAL CONDUCTIVITY HUM HUMIDITY NH3 AMMONIA SOx OXIDES OF SULFUR VIS VISIBLE LIGHT x TYPE OF TRACING INDICATED BY (ET) ELECTRICAL, (ST) STEAM, (CW) 3(57+(',6&5(7,212)7+('(6,*1(5 DEN DENSITY %RH RELATIVE HUMIDITY NOx OXIDES OF NITROGEN SP GR SPECIFIC GRAVITY VISC VISCOSITY CHILLED WATER, ETC DEWPT DEW POINT IR INFRARED O2 OXYGEN TC THERMAL CONDUCTIVITY DO DISSOLVED OXYGEN LC LIQUID CHROMATOGRAPH OP OPACITY TDL TUNABLE DIODE LASER x GENERIC INSTRUMENT CONNECTION TO PROCESS FLOW x GENERIC INSTRUMENT CONNECTION TO EQUIPMENT FLOW CFR CONSTANT FLOW REGULATOR FLT FLOW RATE OP-E ECCENTRIC PV PITOT VENTURI TUR TURBINE x HEAT (COOL) TRACED GENERIC INSTRUMENT IMPULSE LINE CONE CONE LAM LAMINAR OP-FT FLANGE TAPS SNR SONAR US ULTRASONIC x PROCESS LINE OR EQUIPMENT MAY OR MAY NOT BE TRACED COR CORIOLLIS MAG MAGNETIC OP-MH MULTI-HOLE SON SONIC VENT VENTURI TUBE DOP DOPPLER OP ORIFICE PLATE OP-P PIPE TAPS TAR TARGET VOR VORTEX SHEDDING DSON DOPPLER SONIC OP-CT CORNER TAPS OP-VC VENA CONTRACTA TAPS THER THERMAL WDG WEDGE x HEAT(COOL) TRACED INSTRUMENT FLN FLOW NOZZLE OP-CQ CIRCLE QUADRANT PD POSITIVE DISPLACEMENT TTS TRANSIT TIME SONIC x INSTRUMENT IMPULSE LINE MAY OR MAY NOT BE TRACED PT PITOT TUBE LEVEL CAP CAPACITANCE DP DIFFERENTIAL PRESSURE MS MAGNETOSTRICTIVE SON SONIC d/p DIFFERENTIAL PRESSURE GWR GUIDED WAVE RADAR NUC NUCLEAR US ULTRASONIC LINE SYMBOLS DI DIELECTRIC CONSTANT LSR LASER RADAR RADAR LINE TYPE/SYMBOL DESCRIPTION DISP DISPLACER MAG MAGNETIC RES RESISTANCE x IA MAY BE REPLACED BY PA (PLANT AIR), NS (NITROGEN), OR GS (ANY GAS SUPPLY). PRESSURE IA x INDICATE SUPPLY PRESSURE AS REQUIRED, E.G. PA-70 KPA, NS-150 ABS ABSOLUTE MAN MANOMETER VAC VACUUM PSIG, ETC. AVG AVERAGE P-V PRESSURE-VACUUM x INSTRUMENT ELECTRIC POWER SUPPLY. DRF DRAFT SG STRAIN GAUGE ES x INDICATE VOLTAGE AND TYPE AS REQUIRED, E.G. ES-220 VAC x ES MAY BE REPLACED BY 24 VDC, 120 VAC, ETC. TEMPERATURE x INSTRUMENT HYDRAULIC POWER SUPPLY. BM BI-METAL RTD RESISTANCE TEMP. DETECTOR TCJ THERMOCOUPLE, TYPE J THRM THERMISTOR HS x INDICATE PRESSURE AS REQUIRED, E.G. HS-70 PSIG. IR INFRARED TC THERMOCOUPLE TCK THERMOCOUPLE, TYPE K TMP THERMOPILE x ELECTRONIC OR ELECTRICAL CONTINUOUSLY VARIABLE OR BINARY SIGNAL. RAD RADIATION TCE THERMOCOUPLE, TYPE E TCT THERMOCOUPLE, TYPE T TRAN TRANSISTOR x FUNCTIONAL DIAGRAM BINARY SIGNAL. RP RADIATION PYROMETER x FUNCTIONAL DIAGRAM CONTINUOUSLY VARIABLE SIGNAL. MISCELLANEOUS x ELECTRICAL SCHEMATIC LADDER DIAGRAM SIGNAL AND POWER ANNUNCIATION BURNER, COMBUSTION OTHER POSITION RAILS. ALM ALARM FR FLAME ROD CONC CONCENTRIC PB PUSHBUTTON CAP CAPACITANCE x FILLED THERMAL ELEMENT CAPILLARY TUBE. ANN ANNUNCIATOR IGN IGNITER HOA HAND-OFF-AUTO PC PHOTOCELL EC EDDY CURRENT x FILLED SENSING LINE BETWEEN PRESSURE SEAL AND INSTRUMENT. IR TELEVISION L/R LOCAL/REMOTE SMOKE SMOKE IND INDUCTIVE x GUIDED ELECTROMAGNETIC SIGNAL. UV ULTRA VIOLET MOS MAINTENANCE OVERRIDE SWITCH SYNC SYNCHRONIZATION LAS LASER x GUIDED SONIC SIGNAL. MULTI MULTIVARIABLE TDR TIME DELAY RELAY MAG MAGNETIC x FIBER OPTIC SIGNAL. O/L OVERLOAD TEST TEST MECH MECHANICAL x COMMUNICATION LINK AND SYSTEM BUS, BETWEEN DEVICES AND OX OVERRIDE SWITCH VIBR VIBRATION OPT OPTICAL FUNCTIONS OF A SHARED DISPLAY, SHARED CONTROL SYSTEM. NR NARROW RANGE WR WIDE RANGE RADAR RADAR x DCS, PLC, OR PC COMMUNICATION LINK AND SYSTEM BUS. QUANTITY RADIATION SPEED WEIGHT, FORCE x COMMUNICATION LINK OR BUS CONNECTING TWO OR MORE INDEPENDENT MICROPROCESSORS OR COMPUTER-BASED SYSTEMS. PE PHOTOELECTRIC ALPHA RADIATION ACC ACCELERATION LC LOAD CELL x DCS-TO-DCS, DCS-TO-PLC, PLC-TO-PC, DCS-TO-FIELDBUS, ETC, TOG TOGGLE BETA RADIATION EC EDDY CURRENT SG STRAIN GAUGE CONNECTIONS. Y GAMMA RADIATION PROX PROXIMITY WS WEIGH SCALE x COMMUNICATION LINK AND SYSTEM BUS, BETWEEN DEVICES AND n NEUTRON RADIATION VEL VELOCITY FUNCTIONS OF A FIELDBUS SYSTEM. RAD RADIATION ADSORBED DOSE x LINK FROM AND TO "INTELLIGENT" DEVICES. REM ROENTGEN EQUIVALENT MAN x COMMUNICATION LINK BETWEEN A DEVICE AND A REMOTE CALIBRATION ADJUSTMENT DEVICE OR SYSTEM. EQUIPMENT DESCRIPTIONS (NOTE ) x LINK FROM AND TO 'SMART' DEVICES BOUNDARY IDENTIFICATION MISCELLANEOUS IDENTIFICATIONS AIR COOLER HEAT EXCHANGER PRESSURE VESSEL x MECHANICAL LINK OR CONNECTION. TUBE DP/DT: PSIG/°F DESIGN DUTY: BTU/HR DP/DT: PSIG/°F LIMITS OF LIMITS OF TUBE OP/OT : PSIG/°F DUTY CYCLE: % OP/OT: PSIG/°F UPSTREAM PIPE DOWNSTREAM 2 x PRIMARY LINE DESIGN DUTY: BTU/HR SHELL DP/DT: PSIG/°F SIZE: ID X T-T CLASS OR LINE PIPE CLASS OR NUMBER LINE NUMBER REVISION CLOUD x SECONDARY LINE CONFIGURATION: SHELL OP/OT: PSIG/°F CAPACITY: GAL MOTOR NAME PLATE: HP TUBE DP/DT: PSIG/°F AREA A AREA B (ST) (ST) (ST) (ST) (ST) x STEAM TRACE LINE TUBE OP/OT: PSIG/°F PUMP COMPRESSOR CAPACITY: GPM (ET) (ET) (ET) (ET) x ELECTRICAL TRACE LINE DP/DT: PSIG/°F TANK DUTY CYCLE: % VENDOR PACKAGE REVISION TRIANGLE OP/OT: PSIG/°F DP/DT: PSIG/°F DESIGN HEAD: FT DRAWING CONNECTIONS

                                                                                                                                                                                                                                                     //        //         //         //         //        x PNEUMATIC SIGNAL CAPACITY: FT3/MIN              OP/OT: PSIG/°F                    MOTOR NAME PLATE: HP DUTY CYCLE: %                  SIZE: ID X HEIGHT: FT             DP/DT: PSIG/°F              CONNECTOR             DESTINATION                                                                                                              L         L        L          L        L        x HYDRAULIC SIGNAL.

MOTOR NAME PLATE: HP CAPACITY: GAL I.D. NUMBER PAGE NUMBER HOLD CLOUD CHILLER 2 R x REFRIGERANT FILTERS AHU/ ACU/ FCU CAPACITY: TONS XXXXX XXXX-XX-XXXX-X-XX-XXXX-XXX DP/DT: PSIG/°F AIR FLOW: CFM DUTY CYCLE: % TO/FROM OP/OT: PSIG/°F DUTY CYCLE: % EVAP FLOW: GPM HOLD x OTHER SYSTEMS SIZE: ID X T-T COOLING CAPACITY: BTU/ HR EVAP EWT/ LWT: °F/ °F NON-VENDOR PACKAGE XXXX-XX-XXXX-X-XX-XXXX-XXX XXXXX OR MODULE PACKAGE PARTICLE SIZE: MICRON HEATING CAPACITY: BTU/ HR (kW) COND FLOW: GPM TO/FROM MOTOR NAMEPLATE: HP COND EWT/ LWT: °F/ °F SL OP FAN POWER: kW X E XXXX-XX-XXXX-X-XX-XXXX-XXX XXXXX 7 AIR FLOW: CFM UNIT HEATER TO/FROM PFD CALCULATION NODE PIPE SLOPE DUTY CYCLE: % HEATING CAPACITY: kW CONNECTION NUMBER (NODE IDENTIFIED WITH A NUMBER MOTOR NAMEPLATE: HP (E.G., TO CONDENSER) AND/OR LETTER) Tier 2 1.7-12 Revision 4.1

NuScale Final Safety Analysis Report Drawings and Other Detailed Information Figure 1.7-3f: Piping and Instrumentation Diagram Legends PRIMARY ELEMENTS AUXILIARY AND ACCESSORY DEVICES FLOW ANALYSIS 127(6 GENERIC ORIFICE PLATE TURBINE FLOWMETER / PROPELLER FLOWMETER AW SAMPLE INSERT PROBE, FLANGED / SAMPLE WELL,  $%%5(9,$7,216)520)81&7,21,'(17,),(56 FLANGED 7$%/(21),*85(H6+$//%(86(',) LINE NUMBERING FORMAT 025(7+$121((/(0(177<3($33($5621 QUICK-CHANGE ORIFICE PLATE VORTEX SHEDDING FLOWMETER 7+('5$:,1* GG - CC - DDDD - XXXX - TUVX - KK INSULATION CODE CONCENTRIC CIRCLE ORIFICE PLATE TARGET FLOWMETER

                                                                                                                                                                                                                                                                                    $1$%%5(9,$7,21)5207+()81&7,21 AX                                                                                                PIPING LINE SPECIFICATION NUMBER    ,'(17,),(567$%/(21),*85(H6+28/'

SAMPLE ANALYSIS ACCESSORY, FLANGED

                                                                                                                                                                                                                                                                                      %(86('72,'(17,)<7+((/(0(177<3(

ECCENTRIC CIRCLE ORIFICE PLATE MAGNETIC FLOWMETER UNIQUE IDENTIFIER SYSTEM ABBREVIATION CIRCLE QUADRANT ORIFICE PLATE THERMAL MASS FLOWMETER FLOW MODULE NUMBER NOMINAL DIAMETER FX MULTI-HOLE ORIFICE PLATE POSITIVE DISPLACEMENT FLOWMETER FLOW STRAIGHTENING VANES / FLOW CONDITIONING ELEMENT COMPONENT IDENTIFICATION SYSTEM GENERIC VENTURI TUBE, FLOW NOZZLE, OR CONE METER / FLOW TUBE (NOTE 1) CC - DDDD - EEEE - XXXXY - ZZ ATTACHED COMPONENT MODIFIER VENTURI TUBE WEDGE METER UNIQUE IDENTIFIER/SUFFIX P INSTRUMENT PURGE/FLUSHING FLUID OR DEVICES COMPONENT TYPE FLOW NOZZLE CORIOLLIS FLOWMETER SYSTEM ABBREVIATION PRESSURE FLOW TUBE SONIC FLOWMETER / ULTRASONIC FLOWMETER MODULE NUMBER DIAPHRAGM PRESSURE SEAL, GENERIC CONNECTION / INTEGRAL ORIFICE PLATE VARIABLE AREA FLOWMETER DIAPHRAGM CHEMICAL SEAL, GENERIC CONNECTION ENGINEERING DRAWING NUMBERING STANDARD PITOT TUBE OPEN CHANNEL WEIR PLATE AAAA - CC - DDDD - Y - NN - XXXX - SYY SHEET NUMBER DIAPHRAGM PRESSURE SEAL, WELDED/ DIAPHRAGM AVERAGING PITOT TUBE OPEN CHANNEL FLUME CHEMICAL SEAL, WELDED UNIQUE IDENTIFIER ANALYSIS LEVEL DRAWING TYPE INTERNALLY MOUNTED BALL FLOAT TEMPERATURE SINGLE ELEMENT SENSING PROBE DISCIPLINE CODE (MAY BE INSTALLED THROUGH TOP OF VESSEL) TW SYSTEM CODE DUAL ELEMENT SENSING PROBE INTERNALLY MOUNTED DISPLACER THERMOWELL, FLANGED (NOTE 2) MODULE UMBER (BUBBLE MAY BE OMITTED IF CONNECTED TO ANOTHER INSTRUMENT) PROJECT IDENTIFIER FIBEROPTIC SENSING PROBE RADIATION SINGLE POINT BURNER ULTRAVIOLET FLAME DETECTOR / TELEVISION FLAME MONITOR RADIATION, MULTI-POINT OR CONTINUOUS FLAME ROD / FLAME DETECTOR PRIMARY ELEMENT (E.G. DIP TUBE) AND STILLING WELL (MAY BE INSTALLED THROUGH SIDE OF VESSEL) PRESSURE (MAY BE INSTALLED WITHOUT STILLING WELL) STRAIN GAGE OR OTHER ELECTRONIC SENSOR (NOTE 2) PE (BUBBLE MAY BE OMITTED IF CONNECTED TO ANOTHER FLOAT WITH GUIDE WIRES INSTRUMENT) (LOCATION READOUT SHOULD BE NOTED: AT GRADE, AT TOP, OR ACCESSIBLE FROM A LADDER) (GUIDE WIRES MAY BE OMITTED) TEMPERATURE INSERT PROBE (MAY BE INSTALLED THROUGH TOP OF VESSEL) GENERIC ELEMENT WITHOUT THERMOWELL (NOTE 2) TE (BUBBLE MAY BE OMITTED IF CONNECTED TO ANOTHER INSTRUMENT) RADAR Tier 2 1.7-13 Revision 4.1

This section addresses interface requirements between the NuScale Power Plant certified design and the site-specific design provided in the combined license (COL) application. Section 1.2 identifies the structures, systems, and components that are included in the certified design. Figure 1.2-1 provides a representation of the overall facility and Figure 1.2-2 provides the general boundaries between the certified design and site-specific design. Table 1.8-1 identifies the interfaces between the NuScale certified design and the site-specific design. There are two types of interface requirements described:

• CDI: Conceptual design information that is provided for the non-certified portion of the plant to facilitate review of the certified design and to confirm the adequacy of identified interface requirements.
• COL: NuScale design assumptions related to site-specific design elements that are the responsibility of the COL applicant. This type of interface is identified as a COL information item.

1 Combined License Information Items Information that must be provided in order to license and operate a site-specific NuScale Power Plant, but is not included in the certified design, is identified throughout the Final Safety Analysis Report as COL information items. Table 1.8-2 lists the COL information items, includes the COL information item text and identifies the section where the information item is located. The COL applicant addresses each COL information item in the section where it is located. 2 Departures Item 1.8-1: A COL applicant that references the NuScale Power Plant design certification will provide a list of departures from the certified design. 2 1.8-1 Revision 4.1

System, Structure, or Component Interface FSAR Type Section urbine Generator Buildings CDI 1.2.2 nnex Building CDI 1.2.2 ooling towers, pump houses, and associated structures, systems, and CDI 1.2.2, omponents (e.g., cooling tower basin, circulating water pumps, cooling 10.4.5 wer fans, chemical treatment building, etc.) ecurity Buildings CDI 1.2.2 entral Utility Building CDI 1.2.2 iesel Generator Buildings CDI 1.2.2 ffsite power transmission system, main switchyard, and transformer area CDI 8.2 uxiliary AC power system CDI 8.3.1 te cooling water system CDI 9.2.7 irculating water system CDI 10.4.5 rounding and lightning protection system CDI 8.3.1 lant exhaust stack CDI 9.4.2 otable and sanitary water systems COL 9.2.4 esin tanks for the condensate polishing system COL 10.4 te drainage system COL N/A aw water system COL 9.2.9 te parameters, geographic and demographic characteristics, COL Table 2.0-1, 2.1, 2.2, eteorological characteristics, nearby industrial, transportation, and military 2.3, 2.4, 2.5, 3.3, 3.4 cilities, hydrologic characteristics, geology, seismology, and geotechnical haracteristics, weather conditions and site topography, flooding te-specific communications COL 9.5.2 urbine generators COL 3.5-1 perational Support Center COL 13.3 2 1.8-2 Revision 4.1

Item No. Description of COL Information Item Section Item 1.1-1: A COL applicant that references the NuScale Power Plant design certification will identify the 1.1 site-specific plant location. Item 1.1-2: A COL applicant that references the NuScale Power Plant design certification will provide the 1.1 schedules for completion of construction and commercial operation of each power module. Item 1.4-1: A COL applicant that references the NuScale Power Plant design certification will identify the 1.4 prime agents or contractors for the construction and operation of the nuclear power plant. Item 1.7-1: A COL applicant that references the NuScale Power Plant design certification will provide site- 1.7 specific diagrams and legends, as applicable. Item 1.7-2: A COL applicant that references the NuScale Power Plant design certification will list additional 1.7 site-specific piping and instrumentation diagrams and legends as applicable. Item 1.8-1: A COL applicant that references the NuScale Power Plant design certification will provide a list of 1.8 departures from the certified design. Item 1.9-1: A COL applicant that references the NuScale Power Plant design certification will review and 1.9 address the conformance with regulatory criteria in effect six months before the docket date of the COL application for the site-specific portions and operational aspects of the facility design. Item 1.10-1: A COL applicant that references the NuScale Power Plant design certification will evaluate the 1.10 potential hazards resulting from construction activities of the new NuScale facility to the safety-related and risk significant structures, systems, and components of existing operating unit(s) and newly constructed operating unit(s) at the co-located site per 10 CFR 52.79(a)(31). The evaluation will include identification of management and administrative controls necessary to eliminate or mitigate the consequences of potential hazards and demonstration that the limiting conditions for operation of an operating unit would not be exceeded. This COL item is not applicable for construction activities (build-out of the facility) at an individual NuScale Power Plant with operating NuScale Power Modules. Item 2.0-1: A COL applicant that references the NuScale Power Plant design certification will demonstrate 2.0 that site-specific characteristics are bounded by the site parameters specified in Table 2.0-1. If site-specific values are not bounded by the values in Table 2.0-1, the COL applicant will demonstrate the acceptability of the site-specific values in the appropriate sections of its combined license application. Item 2.1-1: A COL applicant that references the NuScale Power Plant design certification will describe the 2.1 site geographic and demographic characteristics. Item 2.2-1: A COL applicant that references the NuScale Power Plant design certification will describe 2.2 nearby industrial, transportation, and military facilities. The COL applicant will demonstrate that the design is acceptable for each of these potential hazards, or provide site-specific design alternatives. Item 2.3-1: A COL applicant that references the NuScale Power Plant design certification will describe the 2.3 site-specific meteorological characteristics for Section 2.3.1 through Section 2.3.5, as applicable. Item 2.4-1: A COL applicant that references the NuScale Power Plant design certification will investigate 2.4 and describe the site-specific hydrologic characteristics for Section 2.4.1 through Section 2.4.14, except Section 2.4.8 and Section 2.4.10. Item 2.5-1: A COL applicant that references the NuScale Power Plant design certification will describe the 2.5 site-specific geology, seismology, and geotechnical characteristics for Section 2.5.1 through Section 2.5.5, below. Item 3.2-1: A COL applicant that references the NuScale Power Plant design certification will update Table 3.2 3.2-1 to identify the classification of site-specific structures, systems, and components. Item 3.3-1: A COL applicant that references the NuScale Power Plant design will confirm that nearby 3.3 structures exposed to severe and extreme (tornado and hurricane) wind loads will not collapse and adversely affect the Reactor Building or Seismic Category I portion of the Control Building. Item 3.4-1: A COL applicant that references the NuScale Power plant design certification will confirm the 3.4 final location of structures, systems, and components subject to flood protection and final routing of piping. 2 1.8-3 Revision 4.1

Item No. Description of COL Information Item Section Item 3.4-2: A COL applicant that references the NuScale Power plant design certification will develop the 3.4 on-site program addressing the key points of flood mitigation. The key points to this program include the procedures for mitigating internal flooding events; the equipment list of structures, systems, and components subject to flood protection in each plant area; and providing assurance that the program reliably mitigates flooding to the identified structures, systems, and components. Item 3.4-3: A COL applicant that references the NuScale Power plant design certification will develop an 3.4 inspection and maintenance program to ensure that each water-tight door, penetration seal, or other degradable measure remains capable of performing its intended function. Item 3.4-4: A COL applicant that references the NuScale Power plant design certification will confirm that 3.4 site-specific tanks or water sources are placed in locations where they cannot cause flooding in the Reactor Building or Control Building. Item 3.4-5: A COL applicant that references the NuScale Power Plant design certification will determine the 3.4 extent of waterproofing and dampproofing needed for the underground portion of the Reactor Building and Control Building based on site-specific conditions. Additionally, a COL applicant will provide the specified design life for waterstops, waterproofing, damp proofing, and watertight seals. If the design life is less than the operating life of the plant, the COL applicant will describe how continued protection will be ensured. Item 3.4-6: A COL applicant that references the NuScale Power Plant design certification will confirm that 3.4 nearby structures exposed to external flooding will not collapse and adversely affect the Reactor Building or Seismic Category I portion of the Control Building. Item 3.4-7: A COL applicant that references the NuScale Power Plant design certification will determine the 3.4 extent of waterproofing and damp proofing needed to prevent groundwater and foreign material intrusion into the expansion gap between the end of the tunnel between the Reactor Building and the Control Building, and the corresponding Reactor Building connecting walls. Item 3.5-1: A COL applicant that references the NuScale Power Plant design certification will demonstrate 3.5 that the site-specific turbine missile parameters are bounded by the design certification analysis, or provide a missile analysis using the site-specific turbine generator parameters to demonstrate that barriers adequately protect essential structures, systems, and components from turbine missiles. Parameters to verify are limiting turbine missile spectrum (rotor and blade material properties); turbine rotor design, geometry and number of blades; final design of the reactor building exterior wall; final design of the control building exterior wall and grade-level slab; and location of the turbines with respect to the reactor building and control building. Item 3.5-2: A COL applicant that references the NuScale Power Plant design certification will address the 3.5 effect of turbine missiles from nearby or co-located facilities. Item 3.5-3: A COL applicant that references the NuScale Power Plant design certification will confirm that 3.5 automobile missiles cannot be generated within a 0.5-mile radius of safety-related structures, systems, and components and risk-significant structures, systems, and components requiring missile protection that would lead to impact higher than 30 feet above plant grade. Additionally, if automobile missiles impact at higher than 30 feet above plant grade, the COL applicant will evaluate and show that the missiles will not compromise safety-related and risk-significant structures, systems, and components. Item 3.5-4: A COL applicant that references the NuScale Power Plant design certification will evaluate site- 3.5 specific hazards for external events that may produce more energetic missiles than the design basis missiles defined in Tier 2, Section 3.5.1.4. Item 3.6-1: A COL applicant that references the NuScale Power Plant design certification will complete the 3.6 routing of piping systems outside of the containment vessel and the area under the bioshield, identify the location of high- and moderate-energy lines, and update Table 3.6-1 as necessary. This activity includes the performance of associated final piping stress analyses, design and qualification of associated piping supports, evaluation of subcompartment pressurization effects (if applicable), and completion of the Balance of Plant Pipe Rupture Hazards Analysis, including the design and evaluation of pipe whip/jet impingement mitigation devices as required. This includes an evaluation and disposition of multi-module impacts in common pipe galleries. 2 1.8-4 Revision 4.1

Item No. Description of COL Information Item Section Item 3.6-2: A COL applicant that references the NuScale Power Plant design certification will verify that the 3.6 pipe rupture hazards analysis (including dynamic and environmental effects) of the high- and moderate-energy lines outside the containment vessel (under the bioshield) is applicable. If changes are required, the COL applicant will update the pipe rupture hazards analysis, design additional protection features as necessary, and update Table 3.6-2. Item 3.6-3: A COL applicant that references the NuScale Power Plant design certification will perform the 3.6 pipe rupture hazards analysis (including dynamic and environmental effects) of the high- and moderate-energy lines outside the reactor pool bay in the Reactor Building (RXB) and update Table 3.6-2 as appropriate. This includes an evaluation and disposition of multi-module impacts in common pipe galleries, and evaluations regarding subcompartment pressurization. The COL applicant will show that the analysis of RXB piping bounds the possible effects of ruptures for the routings of lines outside of the RXB or perform the pipe rupture hazards analysis of the high-and moderate-energy lines outside the buildings. Item 3.6-4: Not used. 3.6 Item 3.7-1: A COL applicant that references the NuScale Power Plant design certification will describe the 3.7 site-specific structures, systems, and components. Item 3.7-2: A COL applicant that references the NuScale Power Plant design certification will provide 3.7 site-specific time histories. In addition to the above criteria for cross correlation coefficients, time step and earthquake duration, strong motion durations, comparison to response spectra and power spectra density, the applicant will also confirm that site-specific ratios V/A and AD/V2 (A, V, D, are peak ground acceleration, ground velocity, and ground displacement, respectively) are consistent with characteristic values for the magnitude and distance of the appropriate controlling events defining the site-specific uniform hazard response spectra. Item 3.7-3: A COL applicant that references the NuScale Power Plant design certification will: 3.7

• develop a site-specific strain compatible soil profile.
• confirm that the criterion for the minimum required response spectrum has been satisfied.
• determine whether the seismic site characteristics fall within the seismic design parameters such as soil layering assumptions used in the certified design, range of soil parameters, shear wave velocity values, and minimum soil bearing capacity.

Item 3.7-4: A COL applicant that references the NuScale Power Plant design certification will confirm that 3.7 nearby structures exposed to a site-specific safe shutdown earthquake will not collapse and adversely affect the Reactor Building or Seismic Category I portion of the Control Building. Item 3.7-5: A COL applicant that references the NuScale Power Plant design certification will perform a 3.7 soil-structure interaction analysis of the Reactor Building and the Control Building using the NuScale SASSI2010 models for those structures. The COL applicant will confirm that the site-specific seismic demands of the standard design for critical structures, systems, and components in Appendix 3B are bounded by the corresponding design certified seismic demands and, if not, the standard design for critical structures, systems, and components will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands. Seismic demands investigated shall include forces, moments, deformations, in-structure response spectra, and seismic stability of the structures. Item 3.7-6: A COL applicant that references the NuScale Power Plant design certification will perform a 3.7 structure-soil-structure interaction analysis that includes the Reactor Building, Control Building, Radioactive Waste Building and both Turbine Generator Buildings. The COL applicant will confirm that the site-specific seismic demands of the standard design structures, systems, and components are bounded by the corresponding design certified seismic demands and, if not, the standard design structures, systems, and components will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands. Item 3.7-7: A COL applicant that references the NuScale Power Plant design certification will provide a 3.7 seismic monitoring system and a seismic monitoring program that satisfies Regulatory Guide 1.12 Nuclear Power Plant Instrumentation for Earthquakes, Rev. 2 (or later) and Regulatory Guide 1.166 Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-earthquake Actions, Rev. 0 (or later). This information is to be provided as noted below. 2 1.8-5 Revision 4.1

Item No. Description of COL Information Item Section Item 3.7-8: A COL applicant that references the NuScale Power Plant design certification will identify the 3.7 implementation milestone for the seismic monitoring program. In addition, a COL applicant that references the NuScale Power Plant design certification will prepare site-specific procedures for activities following an earthquake. These procedures and the data from the seismic instrumentation system will provide sufficient information to determine if the level of earthquake ground motion requiring shutdown has been exceeded. An activity of the procedures will be to address measurement of the post-seismic event gaps between the fuel racks and the pool walls and between the individual fuel racks and to take appropriate corrective action if needed (such as repositioning the racks or assuring that the as-found condition of the racks is acceptable based on the assumptions of the racks' design basis analysis). Acceptable guidance for procedure development is contained in Regulatory Guide 1.166 "Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-earthquake Actions," Rev. 0 (or later) and 1.167, "Restart of a Nuclear Power Plant Shut Down by a Seismic Event," Rev. 0 (or later). Item 3.7-9: A COL applicant that references the NuScale Power Plant design certification will include an 3.7 analysis of performance-based response spectra established at the surface and intermediate depth(s) that take into account the complexities of the subsurface layer profiles of the site and provide a technical justification for the adequacy of vertical to horizontal (V/H) spectral ratios used in establishing the site-specific foundation input response spectra and performance-based response spectra for the vertical direction. Item 3.7-10: A COL applicant that references the NuScale Power Plant design certification will perform a 3.7 site-specific configuration analysis that includes the Reactor Building with applicable configuration layout of the desired NuScale Power Modules. The COL applicant will confirm the following are bounded by the corresponding design certified seismic demands:

1) The in-structure response spectra of the standard design at the foundation and roof. See FSAR Figure 3.7.2-107 and Figure 3.7.2-108 for foundation in-structure response spectra and Figure 3.7.2-113 for roof in-structure response spectra.

2) The maximum forces in the NuScale Power Module lug restraints and skirts. See Table 3B-28.

3) The site-specific in-structure response spectra for the NuScale Power Module at the skirt support will be shown to be bounded by the in-structure response spectra in Figure 3.7.2-156 and Figure 3.7.2-157. The site-specific in-structure response spectra for the NuScale Power Module at the lug restraints will be shown to be bounded by the in-structure response spectra in Figure 3.7.2-158 through Figure 3.7.2-163.

4) The maximum forces and moments in the west wing wall and pool wall. See Table 3B-22b and Table 3B-23b.

5) Not used.

6) The site-specific in-structure response spectra shown immediately below will be shown to be bounded by their corresponding certified in-structure response spectra:

• Reactor Building north exterior wall at EL 75-0: bounded by in-structure response spectra in Figure 3.7.2-110
• Reactor Building west exterior wall at EL 126-0: bounded by in-structure response spectra in Figure 3.7.2-112
• Reactor Building crane wheels at EL 145-6: bounded by in-structure response spectra in Figure 3.7.2-114
• Control Building east wall at EL 76-6: bounded by in-structure response spectra in Figure 3.7.2-119a and Figure 3.7.2-119b
• Control Building south wall at EL 120-0: bounded by in-structure response spectra in Figure 3.7.2-121a and Figure 3.7.2-121b If not, the standard design will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands.

Item 3.7-11: A COL applicant that references the NuScale Power Plant design certification will perform a 3.7 site-specific analysis that assesses the effects of soil separation. The COL applicant will confirm that the in-structure response spectra in the soil separation cases are bounded by the in-structure response spectra shown in FSAR Figure 3.7.2-107 through Figure 3.7.2-122. 2 1.8-6 Revision 4.1

Item No. Description of COL Information Item Section Item 3.7-12: A COL applicant that references the NuScale Power Plant design certification will perform an 3.7 analysis that uses site-specific soil and time histories to confirm the adequacy of the fluid-structure interaction correction factor. Item 3.7-13: A COL applicant that references the NuScale Power Plant design certification will perform a site- 3.7 specific analysis that assesses the effects of non-vertically propagating seismic waves on the free-field ground motions and seismic responses of Seismic Category I structures, systems, and components. Item 3.7-14: A COL applicant that references the NuScale Power Plant design certification will demonstrate 3.7 that the site-specific seismic demand is bounded by the FSAR capacity for an empty dry dock condition. Item 3.7-15: A COL applicant that references the NuScale Power Plant design certification will determine the 3.7 appropriate site-specific number of interaction planes for soil structure interaction. Item 3.7-16: A COL applicant that references the NuScale Power Plant design certification will determine the 3.7 means and methods of lifting the bioshield. A COL applicant will demonstrate that bioshield components and connections can withstand the bioshield loads and appropriate load factors. Item 3.8-1: A COL applicant that references the NuScale Power Plant design certification will describe the 3.8 site-specific program for monitoring and maintenance of the Seismic Category I structures in accordance with the requirements of 10 CFR 50.65 as discussed in Regulatory Guide 1.160. Monitoring is to include below grade walls, groundwater chemistry if needed, base settlements and differential displacements. Item 3.8-2: A COL applicant that references the NuScale Power Plant design certification will confirm that 3.8 the site-independent Reactor Building and Control Building are acceptable for use at the designated site. Item 3.8-3: A COL applicant that references the NuScale Power Plant design certification will identify local 3.8 stiff and soft spots in the foundation soil and address these in the design, as necessary. Item 3.8-4: A COL applicant that references the NuScale Power Plant design certification will evaluate and 3.8 document construction aid elements such as steel beams, Q-decking, formwork, lugs, and other items that are left in place after construction, but that were not part of the certified design, to verify the construction aid elements do not have an appreciable adverse effect on overall mass, stiffness, and seismic demands of the certified building structure. The COL applicant will confirm that these left-in-place construction aid elements will not have adverse effects on safety-related structures, systems, and components per Section 3.7.2. Item 3.8-5: A COL applicant that references the NuScale Power Plant design certification will verify that the 3.8 reactor flange tool (RFT) and embed plates are evaluated using site-specific seismic analysis, and generate seismic loads to the reactor pressure vessel and fuel assemblies that are bounded by the certified design. The design of the structural members will be confirmed by assessing demand-to-capacity ratios for the load combinations in Table 3.8.4-23. The design of the embed plates will be confirmed by assessing demand-to-capacity ratios for the load combinations in Table 3.8.4-1 and Table 3.8.4-2, and applicable design codes in Table 3.8.4-12. In addition, the core plate in-structure response spectra for the RFT location shown in Figure B-34 through Figure B-39 of TR-0916-51502 (NuScale Power Module Seismic Analysis) shall be confirmed against the site specific spectra. If either the demands on the structural members or the embed plates exceed their capacity, or core plate motions do not maintain justifiable margin to limits for the fuel assembly, the COL applicant will address and augment the design per the criteria specified in FSAR Section 3.8.4, and the fuel assembly-imposed load limitations. Item 3.8-6: A COL applicant that references the NuScale Power Plant design certification will verify that the 3.8 construction loads applied to the pool liner plate and its support structure do not exceed 600 psf per American Concrete Institute (ACI)-347, Guide to Formwork for Concrete. Item 3.9-1: A COL applicant that references the NuScale Power Plant design certification will provide the 3.9 applicable test procedures before the start of testing and will submit the test and inspection results from the comprehensive vibration assessment program for the NuScale Power Module, in accordance with Regulatory Guide 1.20. 2 1.8-7 Revision 4.1

Item No. Description of COL Information Item Section Item 3.9-2: A COL applicant that references the NuScale Power Plant design certification will develop 3.9 design specifications and design reports in accordance with the requirements outlined under American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III (Reference 3.9-1). A COL applicant will address any known issues through the reactor vessel internals reliability programs (i.e. Comprehensive Vibration Assessment Program, steam generator programs, etc.) in regards to known aging degradation mechanisms such as those addressed in Section 4.5.2.1. Item 3.9-3: A COL applicant that references the NuScale Power Plant design certification will provide a 3.9 summary of reactor core support structure American Society of Mechanical Engineers (ASME) service level stresses, deformation, and cumulative usage factor values for each component and each operating condition in conformance with ASME Boiler and Pressure Vessel Code Section III Subsection NG. Item 3.9-4: A COL applicant that references the NuScale Power Plant design certification will submit a 3.9 Preservice Testing program for valves as required by 10 CFR 50.55a. Item 3.9-5: A COL applicant that references the NuScale Power Plant design certification will establish an 3.9 Inservice Testing program in accordance with American Society of Mechanical Engineers Operation and Maintenance Code and 10 CFR 50.55a. Item 3.9-6: A COL applicant that references the NuScale Power Plant design certification will identify any 3.9 site-specific valves, implementation milestones, and the applicable American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code (and ASME OM Code Cases) for the preservice and inservice testing programs. These programs are to be consistent with the requirements in the latest edition and addenda of the OM Code incorporated by reference in 10 CFR 50.55a in accordance with the time period specified in 10 CFR 50.55a before the scheduled initial fuel load (or the optional ASME Code Cases listed in Regulatory Guide 1.192 incorporated by reference in 10 CFR 50.55a). Item 3.9-7: Not used. 3.9 Item 3.9-8: A COL applicant that references the NuScale Power Plant design certification will develop 3.9 specific test procedures to allow detection and monitoring of power-operated valve assembly performance sufficient to satisfy periodic verification design basis capability requirements. Item 3.9-9: A COL applicant that references the NuScale Power Plant design certification will develop 3.9 specific test procedures to allow detection and monitoring of emergency core cooling system valve assembly performance sufficient to satisfy periodic verification of design basis capability requirements. Item 3.9-10: A COL applicant that references the NuScale Power Plant design certification will verify that 3.9 evaluations are performed during the detailed design of the main steam lines, using acoustic resonance screening criteria and additional calculations as necessary (e.g., Strouhal number) to determine if there is a concern. The methodology contained in NuScale Comprehensive Vibration Assessment Program Technical Report, TR-0716-50439 is acceptable for this purpose. The COL applicant will update Section 3.9.2.1.1.3 to describe the results of this evaluation. Item 3.9-11: A COL applicant that references the NuScale Power Plant design certification will implement a 3.9 control rod drive system Operability Assurance Program that meets the requirements described in Section 3.9.4.4 and provide a summary of the testing program and results. Item 3.9-12: A COL applicant that references the NuScale Power Plant design certification will perform a 3.9 site-specific seismic analysis in accordance with Section 3.7.2.16. In addition to the requirements of Section 3.7, for sites where the high frequency portion of the site-specific spectrum is not bounded by the control rod drive system, the standard design of NuScale Power Module components will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demand. 2 1.8-8 Revision 4.1

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued) Item No. Description of COL Information Item Section COL Item 3.9-13: A COL applicant that references the NuScale Power Plant design certification will complete an 3.9 assessment of piping systems inside the reactor building to determine the portions of piping to be tested for vibration and thermal expansion. The piping systems within the scope of this testing include American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III, Class 1, 2, and 3 piping systems, other high-energy piping systems inside Seismic Category I structures or those whose failure would reduce the functioning of any Seismic Category I plant feature to an unacceptable level, and Seismic Category I portions of moderate-energy piping systems located outside of containment. The COL applicant may select the portions of piping in the NuScale design for which vibration testing is performed while considering the piping system design and analysis, including the vibration screening and analysis results and scope of testing as identified by the Comprehensive Vibration Assessment Program. COL Item 3.9-14: A COL applicant that references the NuScale Power Plant design certification will develop an 3.9 evaluation methodology for the analysis of secondary-side instabilities in the steam generator design. This methodology will address the identification of potential density wave oscillations in the steam generator tubes, and qualification of the applicable portions of the reactor coolant system integral reactor pressure vessel and steam generator given the occurrence of density wave oscillations, including the effects of reverse fluid flows within the tubes. COL Item 3.10-1: A COL applicant that references the NuScale Power Plant design certification will develop and 3.10 maintain a site-specific seismic and dynamic qualification program. COL Item 3.10-2: A COL applicant that references the NuScale Power Plant design certification will develop the 3.10 equipment qualification database and ensure equipment qualification record files are created for the structures, systems, and components that require seismic qualification. COL Item 3.10-3: A COL applicant that references the NuScale Power Plant design certification will submit an 3.10 implementation program for Nuclear Regulatory Commission approval prior to the installation of the equipment that requires seismic qualification. COL Item 3.11-1: A COL applicant that references the NuScale Power Plant design certification will submit a full 3.11 description of the environmental qualification program and milestones and completion dates for program implementation. COL Item 3.11-2: A COL applicant that references the NuScale Power Plant design certification will develop the 3.11 equipment qualification database and ensure equipment qualification record files are created for the structures, systems, and components that require environmental qualification. COL Item 3.11-3: A COL applicant that references the NuScale Power Plant design certification will implement an 3.11 equipment qualification operational program that incorporates the aspects in Section 3.11-7 specific to the environmental qualification of mechanical and electrical equipment. This program will include an update to Table 3.11-1 to include commodities that support equipment listed in Table 3.11-1. COL Item 3.11-4: A COL applicant that references the NuScale Power Plant design certification will ensure the 3.11 environmental qualification program cited in COL Item 3.11-1 includes a description of how equipment located in harsh conditions will be monitored and managed throughout plant life. This description will include methodology to ensure equipment located in harsh environments will remain qualified if the measured dose is higher than the calculated dose. COL Item 3.12-1: A COL applicant that references the NuScale Power Plant design certification may use a piping 3.12 analysis program other than the programs listed in Section 3.12.4.1; however, the applicant will implement a benchmark program using the models for the NuScale Power Plant standard design. COL Item 3.12-2: A COL applicant that references the NuScale Power Plant design certification will confirm that 3.12 the site-specific seismic response is within the parameters specified in Section 3.7. A COL applicant may perform a site-specific piping stress analysis in accordance with the methodologies described in this section, as appropriate. COL Item 3.13-1: A COL applicant that references the NuScale Power Plant design certification will provide an 3.13 inservice inspection program for American Society of Mechanical Engineers (ASME) Class 1, 2, and 3 threaded fasteners. The program will identify the applicable edition and addenda of ASME Boiler and Pressure Vessel Code Section XI and ensure compliance with 10 CFR 50.55a. Tier 2 1.8-9 Revision 4.1

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued) Item No. Description of COL Information Item Section COL Item 4.2-1 A COL applicant that references the NuScale Power Plant design certification and wishes to 4.2 utilize non-baseload operations will provide justification for the fuel performance codes and methods corresponding to the desired operation. COL Item 5.2-1: Not used. 5.2 COL Item 5.2-2: A COL applicant that references the NuScale Power Plant design certification will provide a 5.2 certified Overpressure Protection Report in compliance with American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III Subarticles NB-7200 and NC-7200 to demonstrate the reactor coolant pressure boundary and secondary system are designed with adequate overpressure protection features, including low temperature overpressure protection features. COL Item 5.2-3: Not used. 5.2 COL Item 5.2-4: A COL applicant that references the NuScale Power Plant design certification will develop and 5.2 implement a Strategic Water Chemistry Plan. The Strategic Water Chemistry Plan will provide the optimization strategy for maintaining primary coolant chemistry and provide the basis for requirements for sampling and analysis frequencies, and corrective actions for control of primary water chemistry consistent with the latest version of the Electric Power Research Institute Pressurized Water Reactor Primary Water Chemistry Guidelines. COL Item 5.2-5: A COL applicant that references the NuScale Power Plant design certification will develop and 5.2 implement a Boric Acid Control Program that includes: inspection elements to ensure the integrity of the reactor coolant pressure boundary components for subsequent service, monitoring of the containment atmosphere for evidence of reactor coolant system leakage, the type of visual or other nondestructive inspections to be performed, and the required inspection frequency. COL Item 5.2-6: A COL applicant that references the NuScale Power Plant design certification will develop 5.2 site-specific preservice examination, inservice inspection, and inservice testing program plans in accordance with Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code and will establish implementation milestones. If applicable, a COL applicant that references the NuScale Power Plant design certification will identify the implementation milestone for the augmented inservice inspection program. The COL applicant will identify the applicable edition of the American Society of Mechanical Engineers Code utilized in the program plans consistent with the requirements of 10 CFR 50.55a. COL Item 5.2-7: A COL applicant that references the NuScale Power Plant design certification will establish 5.2 plant-specific procedures that specify operator actions for identifying, monitoring, and trending reactor coolant system leakage in response to prolonged low leakage conditions that exist above normal leakage rates and below the technical specification limits. The objective of the methods of detecting and trending the reactor coolant pressure boundary leak will be to provide the operator sufficient time to take actions before the plant technical specification limits are reached. COL Item 5.3-1: A COL applicant that references the NuScale Power Plant design certification will establish 5.3 measures to control the onsite cleaning of the reactor pressure vessel during construction in accordance with Regulatory Guide 1.28. COL Item 5.3-2: A COL applicant that references the NuScale Power Plant design certification will develop 5.3 operating procedures to ensure that transients will not be more severe than those for which the reactor design adequacy had been demonstrated. COL Item 5.3-3 A COL applicant that references the NuScale Power Plant design certification will describe their 5.3 reactor vessel material surveillance program consistent with NUREG 0800, Section 5.3.1. Tier 2 1.8-10 Revision 4.1

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued) Item No. Description of COL Information Item Section COL Item 5.4-1: A COL applicant that references the NuScale Power Plant design certification will develop and 5.4 implement a Steam Generator Program for periodic monitoring of the degradation of steam generator components to ensure that steam generator tube integrity is maintained. The Steam Generator Program will be based on the latest revision of Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines, and applicable Electric Power Research Institute steam generator guidelines at the time of the COL application. The elements of the program will include: assessment of degradation, tube inspection requirements, tube integrity assessment, tube plugging, primary-to-secondary leakage monitoring, shell side integrity assessment, primary and secondary side water chemistry control, foreign material exclusion, loose parts management, contractor oversight, self-assessment, and reporting. COL Item 6.2-1: A COL applicant that references the NuScale Power Plant design certification will develop a 6.2 containment leakage rate testing program that will identify which option is to be implemented under 10 CFR 50, Appendix J. Option A defines a prescriptive-based testing approach whereas Option B defines a performance-based testing program. COL Item 6.2-2: A COL applicant that references the NuScale Power Plant design certification will verify that the 6.2 final design of the containment vessel meets the design basis requirement to maintain flange contact pressure at accident temperature, concurrent with peak accident pressure. COL Item 6.2-3: A COL applicant that references the NuScale Power Plant design certification will perform an 6.2 analysis that, in consideration of the as-built containment internal free volume, demonstrates that containment design pressure and temperature bounds containment peak accident pressure and temperature. The evaluation value for containment internal free volume must include margin to address the complex shapes of internal structures and components and manufacturing processes. COL Item 6.3-1: A COL applicant that references the NuScale Power Plant design certification will describe a 6.3 containment cleanliness program that limits debris within containment. The program should contain the following elements:

• Foreign material exclusion controls to limit the introduction of foreign material and debris sources into containment.
• Maintenance activity controls, including temporary changes, that confirm the emergency core cooling system function is not reduced by changes to analytical inputs or assumptions or other activities that could introduce debris or potential debris sources into containment.
• Controls that limit the introduction of coating materials into containment.
• An inspection program to confirm containment vessel cleanliness prior to closing for normal power operation.

COL Item 6.4-1: A COL applicant that references the NuScale Power Plant design certification will comply with 6.4 Regulatory Guide 1.78 Revision 1, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release. COL Item 6.4-2: Not used. 6.4 COL Item 6.4-3: Not used. 6.4 COL Item 6.4-4: Not used. 6.4 COL Item 6.4-5: A COL applicant that references the NuScale Power Plant design certification will specify testing 6.4 and inspection requirements for the control room habitability system and control room envelope integrity testing as specified in Table 6.4-4. COL Item 6.6-1: A COL applicant that references the NuScale Power Plant design certification will implement an 6.6 inservice testing program in accordance with 10 CFR 50.55a(f). COL Item 6.6-2: A COL applicant that references the NuScale Power Plant design certification will develop 6.6 preservice inspection and inservice inspection program plans in accordance with Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), and will establish the implementation milestones for the program. The COL applicant will identify the applicable edition of the ASME BPVC used in the program plan consistent with the requirements of 10 CFR 50.55a. The COL applicant will, if needed, address the use of a single inservice inspection program for multiple NuScale Power Modules, including any alternative to the code that may be necessary to implement such an inservice inspection program. Tier 2 1.8-11 Revision 4.1

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued) Item No. Description of COL Information Item Section COL Item 7.0-1: A COL applicant that references the NuScale Power Plant design certification is responsible for 7.0 demonstrating the stability of the NuScale Power Module during normal and power maneuvering operations for closed-loop module control system subsystems that use reactor power as a control input. COL Item 7.2-1: A COL applicant that references the NuScale Power Plant design certification is responsible for 7.2 the implementation of the life cycle processes for the operation phase for the instrumentation and controls systems, as defined in Institute of Electrical and Electronics Engineers (IEEE) Std 1074-2006 and IEEE Std 1012-2004. COL Item 7.2-2: A COL applicant that references the NuScale Power Plant design certification is responsible for 7.2 the implementation of the life cycle processes for the maintenance phase for the instrumentation and controls systems, as defined in Institute of Electrical and Electronics Engineers (IEEE) Std 1074-2006 and IEEE Std 1012-2004. COL Item 7.2-3: The NuScale Digital instrumentation and controls (I&C) Software Configuration Management 7.2 Plan provides guidance for the retirement and removal of a software product from use. A COL applicant that references the NuScale Power Plant design certification is responsible for the implementation of the life cycle processes for the retirement phase for the instrumentation and controls systems, as defined in Institute of Electrical and Electronics Engineers (IEEE) Std 1074-2006 and IEEE Std 1012-2004. The NuScale Digital I&C Software Configuration Management Plan provides guidance for the retirement and removal of a software product from use. COL Item 8.2-1: A COL applicant that references the NuScale Power Plant design certification will describe the 8.2 site-specific switchyard layout and design, including offsite power connections, control and indication, characteristics of circuit breakers and buses, protective relaying, power supplies, lightning and grounding protection equipment, and conformance with General Design Criteria (GDC) 5. COL Item 8.2-2: A COL applicant that references the NuScale Power Plant design certification will describe the 8.2 site-specific offsite power connection and grid stability studies, including the effects of grid contingencies such as the loss of the largest operating unit on the grid, the loss of one NuScale Power Module, and the loss of the full complement of NuScale Power Modules (up to 12). The study will be performed in accordance with the applicable Federal Energy Regulatory Commission, North American Electric Reliability Corporation, and transmission system operator requirements, including communication agreements and protocols. COL Item 8.2-3: A COL applicant that references the NuScale Power Plant design certification will describe the 8.2 testing of the switchyard and the connections to an offsite power system, if provided, consistent with Regulatory Guide 1.68, Revision 4. The testing description will include the details of initial testing associated with degraded offsite power conditions. COL Item 8.3-1: A COL applicant that references the NuScale Power Plant design certification will describe the 8.3 site-specific location, type, and design of the power source to be used as the auxiliary alternating current power system. COL Item 8.3-2: A COL applicant that references the NuScale Power Plant design certification will describe the 8.3 design of the site-specific electrical heat tracing system. COL Item 8.3-3: A COL applicant that references the NuScale Power Plant design certification will describe the 8.3 design of the site-specific plant grounding grid and lightning protection network. COL Item 9.1-1: A COL applicant that references the NuScale Power Plant design certification will develop plant 9.1 programs and procedures for safe operations during handling and storage of new and spent fuel assemblies, including criticality control. COL Item 9.1-2: A COL applicant that references the NuScale Power Plant design certification will demonstrate 9.1 that an NRC-licensed cask can be lowered into the dry dock and used to remove spent fuel assemblies from the plant. COL Item 9.1-3: A COL applicant that references the NuScale Power Plant design certification will develop 9.1 procedures related to the transfer of spent fuel to a transfer cask. COL Item 9.1-4: A COL applicant that references the NuScale Power Plant design certification will provide the 9.1 periodic testing plan for fuel handling equipment. Tier 2 1.8-12 Revision 4.1

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued) Item No. Description of COL Information Item Section COL Item 9.1-5: The COL applicant that references the NuScale Power Plant design certification will describe the 9.1 process for handling and receipt of critical loads including NuScale Power Modules. COL Item 9.1-6: The COL applicant that references the NuScale Power Plant design certification will provide a 9.1 design for a spent fuel cask and handling equipment including procedures and programs for safe handling. COL Item 9.1-7: The COL applicant that references the NuScale Power Plant design certification will provide a 9.1 description of the program governing heavy loads handling. The program should address

• operating and maintenance procedures
• inspection and test plans
• personnel qualifications and operator training
• detailed description of the safe load paths for movement of heavy loads COL Item 9.1-8: A COL applicant that references the NuScale Power Plant design certification will provide a 9.1 structural evaluation of the spent fuel storage racks, and fuel assemblies located in the racks, and confirm the thermal-hydraulic, criticality, and material analysis aspects of the design remain valid. This evaluation is dependent on the vendor-specific spent fuel storage rack design.

COL Item 9.1-9: A COL applicant that references the NuScale Power Plant design certification will provide a 9.1 neutron absorber material qualification report which demonstrates that the neutron absorber material can meet the neutron attenuation and environmental compatibility design functions described in Technical Report TR-0816-49833. The COL applicant will establish procedures to evaluate the neutron attenuation uncertainty associated with the material lot variability and will establish procedures to inspect the as-manufactured material for contamination and manufacturing defects. COL Item 9.2-1: A COL applicant that references the NuScale Power Plant design certification will select the 9.2 appropriate chemicals for the reactor component cooling water system based on site-specific water quality and materials requirements. COL Item 9.2-2: A COL applicant that references the NuScale Power Plant design certification will describe the 9.2 source and pre-treatment methods of potable water for the site, including the use of associated pumps and storage tanks. COL Item 9.2-3: A COL applicant that references the NuScale Power Plant design certification will describe the 9.2 method for sanitary waste storage and disposal, including associated treatment facilities. COL Item 9.2-4: A COL applicant that references the NuScale Power Plant design certification will provide details 9.2 on the prevention of long-term corrosion and organic fouling in the site cooling water system. COL Item 9.2-5: A COL applicant that references the NuScale Power Plant design certification will identify the 9.2 site-specific water source and provide a water treatment system that is capable of producing water that meets the plant water chemistry requirements. COL Item 9.3-1: A COL applicant that references the NuScale Power Plant design certification will submit a 9.3 leakage control program for systems outside containment that contain (or might contain) accident source term radioactive materials following an accident (including systems and components used in post-accident hydrogen and oxygen monitoring of the containment atmosphere). The leakage control program will include an initial test program, a schedule for re-testing these systems, and the actions to be taken for minimizing leakage from such systems to as low as practical. COL Item 9.3-2: Not used. 9.3 COL Item 9.4-1: A COL applicant that references the NuScale Power Plant design certification will specify a 9.4 periodic testing and inspection program for the normal control room heating ventilation and air conditioning system. COL Item 9.4-2: A COL applicant that references the NuScale Power Plant design certification will specify 9.4 periodic testing and inspection requirements for the Reactor Building heating ventilation and air conditioning system in accordance with Regulatory Guide 1.140. COL Item 9.4-3: A COL applicant that references the NuScale Power Plant design certification will specify 9.4 periodic testing and inspection requirements for the Radioactive Waste Building heating ventilation and air conditioning system. Tier 2 1.8-13 Revision 4.1

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued) Item No. Description of COL Information Item Section COL Item 9.4-4: A COL applicant that references the NuScale Power Plant design certification will specify 9.4 periodic testing and inspection requirements for the Turbine Building heating ventilation and air conditioning system. COL Item 9.5-1: A COL applicant that references the NuScale Power Plant design certification will provide a 9.5 description of the offsite communication system, how that system interfaces with the onsite communications system, as well as how continuous communications capability is maintained to ensure effective command and control with onsite and offsite resources during both normal and emergency situations. COL Item 9.5-2: A COL applicant that references the NuScale Power Plant design certification will determine the 9.5 location for the security power equipment within a vital area in accordance with 10 CFR 73.55(e)(9)(vi)(B). COL Item 10.2-1: Not used. 10.2 COL Item 10.2-2: Not used. 10.2 COL Item 10.2-3: Not used. 10.2 COL Item 10.3-1: A COL applicant that references the NuScale Power Plant design certification will provide a 10.3 site-specific chemistry control program based on the latest revision of the Electric Power Research Institute Pressurized Water Reactor Secondary Water Chemistry Guidelines and Nuclear Energy Institute (NEI) 97-06 at the time of the COL application. COL Item 10.3-2: A COL Applicant that references the NuScale Power Plant design certification will provide a 10.3 description of the flow-accelerated corrosion monitoring program for the steam and power conversion systems based on Generic Letter 89-08 and the latest revision of the Electric Power Research Institute NSAC-202L at the time of the COL application. COL Item 10.4-1: A COL applicant that references the NuScale Power Plant design certification will determine the 10.4 size and number of new and spent resin tanks in the condensate polishing system. COL Item 10.4-2: A COL applicant that references the NuScale Power Plant design certification will describe the 10.4 type of fuel supply for the auxiliary boilers. COL Item 10.4-3: A COL applicant that references the NuScale Power Plant design certification will provide a 10.4 secondary water chemistry analysis. This analysis will show that the size, materials, and capacity of the feedwater treatment system equipment and components satisfies the water quality requirements of the secondary water chemistry program described in Section 10.3.5, and that it is compatible with the chemicals used. COL Item 11.2-1: A COL applicant that references the NuScale Power Plant design certification will ensure mobile 11.2 equipment used and connected to plant systems is in accordance with American National Standards Institute / American Nuclear Society (ANSI/ANS)-40.37, Regulatory Guide (RG) 1.143, 10 CFR 20.1406, NRC IE Bulletin 80-10 and 10 CFR 50.34a. COL Item 11.2-2: A COL applicant that references the NuScale Power Plant design certification will calculate doses 11.2 to members of the public using the site-specific parameters, compare those liquid effluent doses to the numerical design objectives of 10 CFR 50, Appendix I, and comply with the requirements of 10 CFR 20.1302 and 40 CFR 190. COL Item 11.2-3: A COL applicant that references the NuScale Power Plant design certification will perform a site- 11.2 specific evaluation of the consequences of an accidental release of radioactive liquid from the pool surge control system storage tank in accordance with NRC Branch Technical Position 11-6. COL Item 11.2-4: A COL applicant that references the NuScale Power Plant design certification will perform a 11.2 site-specific evaluation using the site-specific dilution flow. COL Item 11.2-5: A COL applicant that references the NuScale Power Plant design certification will perform a 11.2 cost-benefit analysis as required by 10 CFR 50.34a and 10 CFR 50, Appendix I, to demonstrate conformance with regulatory requirements. This cost-benefit analysis is to be performed using the guidance of Regulatory Guide 1.110. COL Item 11.3-1: A COL applicant that references the NuScale Power Plant design certification will perform a 11.3 site-specific cost-benefit analysis. Tier 2 1.8-14 Revision 4.1

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued) Item No. Description of COL Information Item Section COL Item 11.3-2: A COL applicant that references the NuScale Power Plant design certification will calculate doses 11.3 to members of the public using the site-specific parameters, compare those gaseous effluent doses to the numerical design objectives of 10 CFR 50, Appendix I, and comply with the requirements of 10 CFR 20.1302 and 40 CFR 190. COL Item 11.3-3: A COL applicant that references the NuScale Power Plant design certification will perform an 11.3 analysis in accordance with Branch Technical Position 11-5 using the site-specific parameters. COL Item 11.4-1: A COL applicant that references the NuScale Power Plant design certification will describe 11.4 mobile equipment used and connected to plant systems in accordance with American National Standards Institute / American Nuclear Society (ANSI/ANS) 40.37, Regulatory Guide 1.143, 10 CFR 20.1406, NRC IE Bulletin 80-10, and 10 CFR 50.34a. COL Item 11.4-2: A COL applicant that references the NuScale Power Plant design certification will develop a 11.4 site-specific process control program following the guidance of Nuclear Energy Institute (NEI) 07-10A (Reference 11.4-3). COL Item 11.5-1: A COL applicant that references the NuScale Power Plant design certification will describe 11.5 site-specific process and effluent monitoring and sampling system components and address the guidance provided in American National Standards Institute (ANSI) N13.1-2011, ANSI N42.18-2004, and Regulatory Guides 1.21, 1.33, and 4.15. COL Item 11.5-2: A COL applicant that references the NuScale Power design certification will develop an offsite 11.5 dose calculation manual (ODCM) that contains a description of the methodology and parameters used for calculation of offsite doses for gaseous and liquid effluents, using the guidance of Nuclear Energy Institute (NEI) 07-09A (Reference 11.5-8). COL Item 11.5-3: A COL applicant that references the NuScale Power design certification will develop a 11.5 Radiological Environmental Monitoring Program (REMP), consistent with the guidance in NUREG-1301 and NUREG-0133, that considers local land use census data for the identification of potential radiation pathways radioactive materials present in liquid and gaseous effluents, and direct external radiation from systems, structures, and components. COL Item 12.1-1: A COL applicant that references the NuScale Power Plant design certification will describe the 12.1 operational program to maintain exposures to ionizing radiation as far below the dose limits as practical, as low as reasonably achievable (ALARA). COL Item 12.2-1: A COL applicant that references the NuScale Power Plant design certification will describe 12.2 additional site-specific contained radiation sources that exceed 100 millicuries (including sources for instrumentation and radiography) not identified in Section 12.2.1. COL Item 12.3-1: A COL applicant that references the NuScale Power Plant design certification will develop the 12.3 administrative controls regarding access to high radiation areas per the guidance of Regulatory Guide 8.38. COL Item 12.3-2: A COL applicant that references the NuScale Power Plant design certification will develop the 12.3 administrative controls regarding access to very high radiation areas per the guidance of Regulatory Guide 8.38. COL Item 12.3-3: A COL applicant that references the NuScale Power Plant design certification will specify 12.3 personnel exposure monitoring hardware, specify contamination identification and removal hardware, and establish administrative controls and procedures to control access into and exiting the radiologically controlled area. COL Item 12.3-4: A COL applicant that references the NuScale Power Plant design certification will develop the 12.3 processes and programs necessary for the implementation of 10 CFR 20.1501 related to conducting radiological surveys, maintaining proper records, calibration of equipment, and personnel dosimetry. COL Item 12.3-5: A COL applicant that references the NuScale Power Plant design certification will describe 12.3 design criteria for locating additional area radiation monitors. COL Item 12.3-6: A COL applicant that references the NuScale Power Plant design certification will develop the 12.3 processes and programs necessary for the use of portable airborne monitoring instrumentation, including accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident. Tier 2 1.8-15 Revision 4.1

NuScale Final Safety Analysis Report Interfaces with Certified Design Table 1.8-2: Combined License Information Items (Continued) Item No. Description of COL Information Item Section COL Item 12.3-7: A COL applicant that references the NuScale Power Plant design certification will develop the 12.3 processes and programs associated with Objectives 5 and 6, to work in conjunction with design features, necessary to demonstrate compliance with 10 CFR 20.1406, and the guidance of Regulatory Guide 4.21. COL Item 12.3-8: A COL applicant that references the NuScale Power Plant design certification will describe the 12.3 radiation shielding design measures used to compensate for major shield wall penetrations in accordance with FSAR Section 12.1.2.3.2 "Minimizing Radiation Levels in Plant Access Areas and Vicinity of Equipment," Section 12.3.1.2.3 "Penetrations," and Section 12.3.2.2 "Design Considerations." Penetration compensatory measures will account for the protection of equipment, and exposures to workers and the public. COL Item 12.4-1: A COL applicant that references the NuScale Power Plant design certification will estimate doses 12.4 to construction personnel from a co-located existing operating nuclear power plant that is not a NuScale Power Plant. COL Item 12.5-1: A COL applicant that references the NuScale Power Plant design certification will describe 12.5 elements of the operational radiation protection program to ensure that occupational and public radiation exposures are as low as reasonably achievable in accordance with 10 CFR 20.1101. COL Item 13.1-1: A COL applicant that references the NuScale Power Plant design certification will provide a 13.1 description of the corporate or home office management and technical support organization, including a description of the qualification requirements for (1) each identified position or class of positions that provide technical support to the onsite operating organization, and (2) individuals holding management and supervisory positions in organizational units providing technical support to the onsite operating organization. COL Item 13.1-2: A COL applicant that references the NuScale Power Plant design certification will provide a 13.1 description of the proposed structure, functions, and responsibilities of the onsite organization necessary to operate and maintain the plant. The proposed operating staff shall be consistent with the minimum licensed operator staffing requirements in Section 18.5. COL Item 13.1-3: A COL applicant that references the NuScale Power Plant design certification will provide a 13.1 description of the qualification requirements for each management, operating, technical, and maintenance position described in the operating organization. COL Item 13.2-1: A COL applicant that references the NuScale Power Plant design certification will provide a 13.2 description and schedule of the initial training and qualification as well as requalification programs for reactor operators and senior reactor operators. COL Item 13.2-2: A COL applicant that references the NuScale Power Plant design certification will provide a 13.2 description and schedule of the non-licensed plant staff training programs including initial training, periodic retraining, and qualification requirements. COL Item 13.3-1: A COL applicant that references the NuScale Power Plant design certification will provide a 13.3 description of the onsite operational support center (OSC) including the direct communication system or systems between the OSC and the control room. COL Item 13.3-2: A COL applicant that references the NuScale Power Plant design certification will provide a 13.3 description of an emergency operations facility for management of overall licensee emergency response. The facility will meet the requirements of 10 CFR 50.47(b)(8) and Section IV.E, "Emergency Facilities and Equipment," of Appendix E to 10 CFR Part 50. COL Item 13.3-3: A COL applicant that references the NuScale Power Plant design certification will provide a 13.3 comprehensive emergency plan in accordance with 10 CFR 50.47, 10 CFR 50, Appendix E, 10 CFR 52.48, and 10 CFR 52.79(a)(21). Tier 2 1.8-16 Revision 4.1

Item No. Description of COL Information Item Section Item 13.4-1: A COL applicant that references the NuScale Power Plant design certification will provide site- 13.4 specific information, including implementation schedule, for operational programs:

• Inservice inspection programs (refer to Section 5.2, Section 5.4, and Section 6.6)
• Inservice testing programs (refer to Section 3.9 and Section 5.2)
• Environmental qualification program (refer to Section 3.11)
• Pre-service inspection program (refer to Section 5.2 and Section 5.4)
• Reactor vessel material surveillance program (refer to Section 5.3)
• Pre-service testing program (refer to Section 3.9.6, Section 5.2, and Section 6.6)
• Containment leakage rate testing program (refer to Section 6.2)
• Fire protection program (refer to Section 9.5)
• Process and effluent monitoring and sampling program (refer to Section 11.5)
• Radiation protection program (refer to Section 12.5)
• Non-licensed plant staff training program (refer to Section 13.2)
• Reactor operator training program (refer to Section 13.2)
• Reactor operator requalification program (refer to Section 13.2)
• Emergency planning (refer to Section 13.3)
• Process control program (PCP) (refer to Section 11.4)
• Security (refer to Section 13.6)
• Quality assurance program (refer to Section 17.5)
• Maintenance rule (refer to Section 17.6)
• Motor-operated valve testing (refer to Section 3.9)
• Initial test program (refer to Section 14.2)

Item 13.5-1: A COL applicant that references the NuScale Power Plant design certification will describe the 13.5 site-specific procedures that provide administrative control for activities that are important for the safe operation of the facility consistent with the guidance provided in Regulatory Guide 1.33, Revision 3. Item 13.5-2: A COL applicant that references the NuScale Power Plant design certification will describe the 13.5 site-specific procedures that operators use in the main control room and locally in the plant, including normal operating procedures, abnormal operating procedures, and emergency operating procedures. The COL applicant will describe the classification system for these procedures, and the general format and content of the different classifications. Item 13.5-3: A COL applicant that references the NuScale Power Plant design certification will describe the 13.5 site-specific maintenance and other operating procedures, including how these procedures are classified, and the general format and content of the different classifications. The categories of procedures listed below should be included:

• plant radiation protection procedures
• emergency preparedness procedures
• calibration and test procedures
• chemical-radiochemical control procedures
• radioactive waste management procedures
• maintenance and modification procedures
• material control procedures
• plant security procedures Item 13.5-4: A COL applicant that references the NuScale Power Plant design certification will provide a plan 13.5 for the development, implementation, and control of administrative procedures, including preliminary schedules for preparation and target dates for completion. Additionally, the COL applicant will identify the group within the operating organization responsible for maintaining these procedures.

2 1.8-17 Revision 4.1

Item No. Description of COL Information Item Section Item 13.5-5: A COL applicant that references the NuScale Power Plant design certification will provide a plan 13.5 for the development, implementation, and control of operating procedures, including preliminary schedules for preparation and target dates for completion. Additionally, the COL applicant will identify the group within the operating organization responsible for maintaining these procedures. Item 13.5-6: Not used. 13.5 Item 13.5-7: A COL applicant that references the NuScale Power Plant design certification will provide a plan 13.5 for the development, implementation, and control of emergency operating procedures (EOPs), including preliminary schedules for preparation and target dates for completion. Included in the submittal is the Procedures Generation Package, consisting of the following:

• Plant-Specific Technical Guidelines, which are guidelines based on analysis of transients and accidents that are specific to the COL applicant's plant design and operating philosophy.
• A plant-specific writers guide that details the specific methods to be used by the COL applicant in preparing EOPs based on the Plant-Specific Technical Guidelines.
• A description of the program for verification and validation of the EOPs.
• A description of the program for training operators on the EOPs.

Additionally, the COL applicant will identify the group within the operating organization responsible for maintaining these procedures. Item 13.5-8: A COL applicant that references the NuScale Power Plant design certification will provide a plan 13.5 for the development, implementation, and control of maintenance and other operating procedures, including preliminary schedules for preparation and target dates for completion. Additionally, the COL applicant will identify what group or groups within the operating organization have the responsibility for maintaining and following these procedures. Item 13.6-1: A COL applicant that references the NuScale Power Plant design certification will provide the 13.6 following:

• Security Plans (Physical Security, Security Training and Qualification, and Safeguards Contingency Plans)
• proposed site security provisions to be implemented during construction and as modules are completed and become operational of a new plant
• portions of the physical security system not located within the nuclear island and structures Item 13.6-2: A COL applicant that references the NuScale Power Plant design certification will be responsible 13.6 for the requirements described in Table 5-1 of TR-0416-48929, NuScale Design of Physical Security Systems (Reference 13.6-1).

Item 13.6-3: A COL applicant that references the NuScale Power Plant design certification will provide a 13.6 secondary alarm station that is equal and redundant to the central alarm station. Item 13.6-4: A COL applicant that references the NuScale Power Plant design certification will provide 13.6 inspections, tests, analyses, and acceptance criteria for site-specific physical security structures, systems, and components. Item 13.6-5: A COL applicant that references the NuScale Power Plant design certification will provide a 13.6 description of the access authorization program. Item 13.6-6: A COL applicant that references the NuScale Power Plant design certification will provide a 13.6 Cyber Security Plan. Item 13.7-1: A COL applicant that references the NuScale Power Plant design certification will provide a 13.7 description of the applicants 10 CFR 26 compliant fitness-for-duty (FFD) program for plant operations. Item 13.7-2: A COL applicant that references the NuScale Power Plant design certification will provide a 13.7 description of the applicants 10 CFR 26 compliant fitness-for-duty (FFD) program for construction. Item 14.2-1: A COL applicant that references the NuScale Power Plant design certification will describe the 14.2 site-specific organizations that manage, supervise, or execute the Initial Test Program, including the associated training requirements. 2 1.8-18 Revision 4.1

Item No. Description of COL Information Item Section Item 14.2-2: A COL applicant that references the NuScale Power Plant design certification is responsible for 14.2 the development of the Startup Administration Manual that will contain the administrative procedures and requirements that control the activities associated with the Initial Test Program. The COL applicant will provide a milestone for completing the Startup Administrative Manual and making it available for NRC inspection. Item 14.2-3: A COL applicant that references the NuScale Power Plant design certification will identify the 14.2 specific operator training to be conducted during low-power testing related to the resolution of TMI Action Plan Item I.G.1, as described in NUREG-0660, NUREG-0694, and NUREG-0737. Item 14.2-4: A COL applicant that references the NuScale Power Plant design certification will provide a 14.2 schedule for the Initial Test Program. Item 14.2-5: A COL applicant that references the NuScale Power Plant design certification will provide a test 14.2 abstract for the potable water system pre-operational testing. Item 14.2-6: A COL applicant that references the NuScale Power Plant design certification will provide a test 14.2 abstract for the seismic monitoring system pre-operational testing. Item 14.2-7: A COL applicant that references the NuScale Power Plant design certification will select the plant 14.2 configuration to perform the Island Mode Test (number of NuScale Power Modules in service). Item 14.3-1: A COL applicant that references the NuScale Power Plant design certification will provide the 14.3 site-specific selection methodology and inspections, tests, analyses, and acceptance criteria for emergency planning. Item 14.3-2: A COL applicant that references the NuScale Power Plant design certification will provide the 14.3 site-specific selection methodology and inspections, tests, analyses, and acceptance criteria for structures, systems, and components within their scope. Item 16.1-1: A COL applicant that references the NuScale Power Plant design certification will provide the 16.1 final plant-specific information identified by [ ] in the generic Technical Specifications and generic Technical Specification Bases. Item 16.1-2 A COL applicant that references the NuScale Power Plant design certification will prepare and 16.1 maintain an owner-controlled requirements manual that includes owner-controlled limits and requirements described in the Bases of the Technical Specifications or as otherwise specified in the FSAR. Item 16.1-3 A COL applicant that references the NuScale Power Plant design certification, and uses 16.1 allocations for sensor response times based on records of tests, vendor test data, or vendor engineering specifications as described in the Bases for Surveillance Requirement 3.3.1.3, will do so for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. Item 17.4-1: A COL applicant that references the NuScale Power Plant design certification will describe the 17.4 Reliability Assurance Program conducted during the operations phases of the plants life. Item 17.4-2: A COL applicant that references the NuScale Power Plant design certification will identify 17.4 site-specific structures, systems, and components within the scope of the Reliability Assurance Program. Item 17.4-3: A COL applicant that references the NuScale Power Plant design certification will identify the 17.4 quality assurance controls for the Reliability Assurance Program structures, systems, and components during site-specific design, procurement, fabrication, construction, and preoperational testing activities. Item 17.5-1: A COL applicant that references the NuScale Power Plant design certification will describe the 17.5 Quality Assurance Program applicable to site-specific design activities and to the construction and operations phases. Item 17.6-1: A COL applicant that references the NuScale Power Plant design certification will describe the 17.6 program for monitoring the effectiveness of maintenance required by 10 CFR 50.65. Item 18.5-1: A COL applicant that references the NuScale Power Plant design certification will address the 18.5 staffing and qualifications of non-licensed operators. 2 1.8-19 Revision 4.1

Item No. Description of COL Information Item Section Item 18.12-1: A COL applicant that references the NuScale Power Plant design certification will provide a 18.12 description of the human performance monitoring program in accordance with applicable NUREG-0711 or equivalent criteria. Item 19.1-1: A COL applicant that references the NuScale Power Plant design certification will identify and 19.1 describe the use of the probabilistic risk assessment in support of licensee programs being implemented during the COL application phase. Item 19.1-2: A COL applicant that references the NuScale Power Plant design certification will identify and 19.1 describe specific risk-informed applications being implemented during the COL application phase. Item 19.1-3: A COL applicant that references the NuScale Power Plant design certification will specify and 19.1 describe the use of the probabilistic risk assessment in support of licensee programs during the construction phase (from issuance of the COL up to initial fuel loading). Item 19.1-4: A COL applicant that references the NuScale Power Plant design certification will specify and 19.1 describe risk-informed applications during the construction phase (from issuance of the COL up to initial fuel loading). Item 19.1-5: A COL applicant that references the NuScale Power Plant design certification will specify and 19.1 describe the use of the probabilistic risk assessment in support of licensee programs during the operational phase (from initial fuel loading through commercial operation). Item 19.1-6: A COL applicant that references the NuScale Power Plant design certification will specify and 19.1 describe risk-informed applications during the operational phase (from initial fuel loading through commercial operation). Item 19.1-7: A COL applicant that references the NuScale Power Plant design certification will evaluate 19.1 site-specific external event hazards (e.g., liquefaction, slope failure), screen those for risk-significance, and evaluate the risk associated with external hazards that are not bounded by the design certification. Item 19.1-8: A COL applicant that references the NuScale Power Plant design certification will confirm the 19.1 validity of the key assumptions and data used in the design certification application probabilistic risk assessment (PRA) and modify, as necessary, for applicability to the as-built, as-operated PRA. Item 19.2-1: A COL applicant that references the NuScale Power Plant design certification will develop severe 19.2 accident management guidelines and other administrative controls to define the response to beyond-design-basis events. Item 19.2-2: A COL applicant that references the NuScale Power Plant design certification will use the 19.2 site-specific probabilistic risk assessment to evaluate and identify improvements in the reliability of core and containment heat removal systems as specified by 10 CFR 50.34(f)(1)(i). Item 19.2-3: A COL applicant that references the NuScale Power Plant design certification will evaluate 19.2 severe accident mitigation design alternatives screened as not required for design certification application. Item 19.3-1: A COL applicant that references the NuScale Power Plant design certification will identify 19.3 site-specific regulatory treatment of nonsafety systems (RTNSS) structures, systems, and components and applicable RTNSS process controls. Item 20.1-1: Not used. 20.1 Item 20.1-2: Not used. 20.1 Item 20.1-3: Not used. 20.1 Item 20.1-4: Not used. 20.1 Item 20.1-5: Not used. 20.1 Item 20.1-6: Not used. 20.1 Item 20.1-7: Not used. 20.1 2 1.8-20 Revision 4.1

Item No. Description of COL Information Item Section Item 20.1-8: A COL applicant that references the NuScale Power Plant design certification will develop 20.1 procedures, training, and a qualification program for operations, maintenance, testing, and calibration of ultimate heat sink level instrumentation to ensure the level instruments will be available when needed and personnel are knowledgeable in interpreting the information as addressed in Nuclear Energy Institute (NEI) 12-02, Revision 1, Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation. Item 20.2-1: A COL applicant that references the NuScale Power Plant design certification will develop 20.2 enhanced firefighting capabilities in accordance with 10 CFR 50.155(b)(2). The enhanced firefighting capabilities should address the expectation elements listed in Section 4.1.3 of the Technical Report TR-0816-50796 (Reference 20.2-1). Item 20.2-2: A COL applicant that references the NuScale Power Plant design certification will provide a 20.2 means for water spray scrubbing using fog nozzles and the availability of water sources, and address runoff water containment issues (sandbags, portable dikes, etc.) as an attenuation measure for mitigating radiation releases outside containment. Item 20.3-1: Not used. 20.3 Item 20.4-1: Not used. 20.4 Item 20.4-2: Not used. 20.4 Item 20.4-3: Not used. 20.4 Item 20.4-4: Not used. 20.4 Item 20.4-5: Not used. 20.4 Item 20.4-6: Not used. 20.4 2 1.8-21 Revision 4.1

This section provides a guide to conformance with regulatory criteria in individual table format, as listed below. Conformance is assessed to regulatory criteria in effect six months before the anticipated docket date. Table 1.9-1, "Conformance Status Legend," defines the codes used to indicate conformance in Table 1.9-2 through Table 1.9-8 Table 1.9-2, Conformance with Regulatory Guides Table 1.9-3, Conformance with NUREG-0800, Standard Review Plan (SRP) and Design Specific Review Standard (DSRS) Table 1.9-4, Conformance with Interim Staff Guidance (ISG) Table 1.9-5, Conformance with TMI Requirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) Table 1.9-6, Evaluation of Operating Experience (Generic Letters and Bulletins) Table 1.9-7, Conformance with Advanced and Evolutionary Light Water Reactor Design Issues (SECYs and associated SRMs) Table 1.9-8, Conformance with SECY-93-087, "Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor Designs" Item 1.9-1: A COL applicant that references the NuScale Power Plant design certification will review and address the conformance with regulatory criteria in effect six months before the submittal date of the COL application for the site-specific portions and operational aspects of the facility design. 1 Conformance with Regulatory Guides Table 1.9-2 provides an evaluation of conformance with the guidance in NRC regulatory guides in effect 6 months before the submittal date of the Final Safety Analysis Report (FSAR). This evaluation also includes an identification and description of deviations from the guidance in the NRC Regulatory Guides as well as suitable justifications for any alternative approaches proposed. The conformance evaluation was performed on the following groups of Regulatory Guides:

• Division 1, Power Reactors
• Division 4, Environmental and Siting (applies to the environmental report and should be discussed therein)
• Division 5, Materials and Plant Protection (applies to the security plan and should be discussed therein)
• Division 8, Occupational Health 2 1.9-1 Revision 4.1

NuScale performed a review of the SRP including Branch Technical Positions and guidance referenced within the SRP. A summary of this review was submitted to the NRC as NP-RT-0612-023, "Gap Analysis Summary Report," Revision 1, in July 2014 (Reference 1.9-1). The gap analysis review for applicability was directed towards the acceptance criteria of each SRP section. However, the review considered the relevance of sub-tier guidance whether referenced in the acceptance criteria or in other portions of the SRP section being reviewed. Additionally, NuScale considered conformance with the DSRS developed by the NRC for the review of the NuScale Power small modular reactor design. This information has been incorporated into Table 1.9-3. Conformance with NRC Interim Staff Guidance is presented in Table 1.9-4. 3 Generic Issues In accordance with 10 CFR 52.47(a)(8), conformance is assessed against technically relevant Three Mile Island (TMI) requirements identified in 10 CFR 50.34(f), except for paragraphs (f)(1)(xii), (f)(2)(ix), and (f)(3)(v). Plant characteristics and plant programs that address relevant TMI requirements are described in the appropriate FSAR sections. In accordance with 10 CFR 52.47(a)(21), proposed resolutions must be identified for any technically relevant unresolved safety issues and medium-priority to high-priority generic safety issues (GSI) identified in the version of NUREG-0933 that is current six months prior to the application for design certification. Resolution and closure of generic issues is managed via the NRC Generic Issues Program. NRC SECY-07-0110, dated July 6, 2007 provides the most recent supplemental status report of the Generic Issues Program prior to the FSAR submittal. As such, Appendix B of NUREG-0933, Rev. 21 (including the Main Report and Supplements 1-34) and NRC letter SECY-07-0110, were used to identify those generic issues applicable to the NuScale Power Plant design certification. Table 1.9-5 identifies the applicable TMI requirements and generic issues, along with an abbreviated summary description of the NRC position for each table entry. Table 1.9-5 also provides a brief conformance assessment notation, including annotation of any exceptions, and a reference to the FSAR section(s) addressing the issue. Those NUREG-0933 generic issues determined as non-applicable were eliminated from consideration in Table 1.9-5 based on these:

• Resolved: Issue has been completely resolved and removed from the latest Generic Issues Program list of Active and Regulatory Office Implementation Generic Issues.
• BWR, Ice Condenser Containment or Other: Issue applies to another nuclear power plant design concept or to the design of a nuclear facility other than a nuclear power plant.

4 Operational Experience (Generic Communications) Per 10 CFR 52.47(a)(22) requirements, applicants for design certification of new plant designs include a description of how operational experience has been incorporated into the design process. Operational experience insights are incorporated into applicable SRP 2 1.9-2 Revision 4.1

application docket date are incorporated into the design unless stated otherwise. The design is an evolution of nuclear power plant designs that have been operated in the United States, as addressed by 10 CFR 52.41(b)(1); hence NRC guidance for technically relevant operational experience issues is addressed in the appropriate FSAR sections. The conformance assessment relative to operational experience is provided in Table 1.9-6, "Evaluation of Operating Experience (Generic Letters and Bulletins)." Further, 10 CFR 21 notifications were reviewed for impact to the NuScale design as part of the supplier evaluation process. NuScales QA supplier evaluation program includes a review of 10 CFR 21 notifications for every Nuclear Safety Related supplier prior to use as an approved supplier for safety related items/services. The evaluation for any 10 CFR 21 notifications is also performed as part of monitoring of supplier performance by periodic annual review. There have been no 10 CFR 21 notifications impacting nuclear safety related work performed by NuScale approved safety related suppliers for the development of the NuScale Design. Therefore, all applicable 10 CFR 21 notifications have been evaluated. 5 Advanced and Evolutionary Light-Water Reactor Design Issues Guidance in SRP Section 1.0 recommends that this section address the applicable licensing and policy issues developed by the NRC and documented in SECY-93-087, "Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs," Secretary of the Commission, Office of the U.S. Nuclear Regulatory Commission, April, 2, 1993, as supplemented by the associated staff requirements memorandum (SRM) dated July 21, 1993. Table 1.9-7 lists applicable design issues identified in SECYs and their associated SRMs. The table provides a conformance assessment notation, including annotation of any exceptions, for each issue. Table 1.9-7 also provides a cross-reference from the SECY issues to the FSAR sections that address them. Table 1.9-8 provides a separate assessment of SECY-93-087 line items pertaining to ALWR designs. 6 References 1.9-1 NuScale Power, LLC, NP-RT-0612-023, Rev 1, Gap Analysis Summary Report. 2 1.9-3 Revision 4.1

formance Status Description Code forms The regulation or regulatory guidance is relevant and applicable, and can be applied as-is. The design fully conforms to the requirement or guidance described in the Section(s) identified. Where options are identified in the regulation or regulatory guidance, Conforms indicates that the design fully conforms to the option(s) selected. ially Conforms The design conforms to those portions of the requirement or guidance that can be appropriately applied as written. The underlying purpose or intent of the requirement or guidance is relevant to the design but cannot be appropriately applied as written, or some portion of the requirement or guidance is applicable while other portions are not applicable. The following are examples:

• A portion of the regulatory requirement or guidance is literally applicable, but the specific language refers to a different type of light water reactor (LWR) design or structures, systems, and components (SSC) that are not part of the design.
• The regulatory requirement or guidance is applicable except for aspects that are specific to combined license or early site permit applicants, or to BWR designs, etc.
• The intent of a regulatory requirement or guidance is applicable, but the specific language refers to one of the following:

                 - a different type of LWR design
                 - an SSC that is not part of the design, but for which a substantively equivalent function is served by other SSC within the design Applicable      The regulation or guidance is not appropriate to apply and therefore conformance is not required.

The following are examples:

• The regulatory requirement or guidance is applicable only to BWR designs.
• The regulatory requirement or guidance is applicable only to large pressurized water reactor (PWR) designs.
• The regulatory requirement or guidance is applicable to the design, but is the responsibility of the COL applicant.
• The regulatory requirement or guidance is applicable to SSC that are not part of the design.

arture For items found within the Code of Federal Regulations (CFR): the regulation is literally applicable; however NuScale intends to depart from the regulation based on the design or safety basis of the NuScale design. That is to say that conformance to the regulation would have a minimal, or even negative, impact on safety of the NuScale design and hence a departure from the regulation (that was originally created for traditional LWRs) is warranted. The form of the departure may be through an exemption request under 10 CFR 52.7 or through a specific process available for a set of regulations. For example, the introduction to 10 CFR 50, Appendix A provides a departure from the General Design Criteria as explained in Section 3.1, and the TMI action items may be identified and justified as not technically relevant to the design consistent with 10 CFR 52.47(a)(8) and 10 CFR 50.34(f).1 1 Note that some TMI action items are categorized as Partially Conforms or Not Applicable rather than Departure. The difference is that those requirements are not applicable by their own terms, for example because they apply to BWRs or to an SSC that the NuScale design lacks. A departure from a TMI requirement is appropriate where the requirement is literally applicable but is inappropriate to apply to the NuScale design. 2 1.9-4 Revision 4.1

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Assumptions Used for Evaluating 2 Not Applicable This guidance is only applicable to BWRs. Not Applicable the Potential Radiological Conse-quences of a Loss of Coolant Acci-dent for Boiling Water Reactors Assumptions Used for Evaluating 2 Not Applicable This RG pertains to existing reactors; RG 1.183 is Not Applicable the Potential Radiological Conse- specified in SRP Section 15.0.3 to be used for new quences of a Loss of Coolant Acci- reactors. dent for Pressurized Water Reactors Safety Guide 5 - Assumptions - Not Applicable This guidance is only applicable to BWRs. Not Applicable Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors Safety Guide 6 - Independence - Partially Conforms The onsite electrical AC power systems do not con- 8.3 Between Redundant Standby tain Class 1E distribution systems. The EDSS design (Onsite) Power Sources and conforms to the guidance for independence of Between Their Distribution Sys- standby power sources and their distribution sys-tems tems. Control of Combustible Gas Con- 3 Partially Conforms The NuScale design complies with the intent of RG 6.2.5 centrations in Containment 1.7 regulatory positions that address hydrogen and oxygen monitors, atmosphere mixing, hydrogen gas production, and containment structural integrity. However, the NuScale design differs from the system designs the guidance addresses. The NuScale design Conformance with Regulatory Criteria combustible gas control system does not use com-bustible gas control systems. The NuScale design supports an exemption to 10 CFR 50.44(c)(2) as described in DCA Part 7, section 2. Qualification and Training of Per- 3 Not Applicable Site-specific programmatic and operational activi- Not Applicable sonnel for Nuclear Power Plants ties are the responsibility of the COL applicant. Application and Testing of Safety- 4 Not Applicable The NuScale design does not require or include Not Applicable Related Diesel Generators in safety-related emergency diesel generators. Nuclear Power Plants Instrument Lines Penetrating the 1 Not Applicable No lines penetrate the NPM containment. Not Applicable Primary Reactor Containment

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Nuclear Power Plant Instrumenta- 2 Partially Conforms Selection of specific equipment is the responsibility 3.7 tion for Earthquakes of the COL applicant or licensee. In addition, seismic 12.3 detectors cannot be installed inside the contain-ment so Section 3.7.3 indicates they are installed in the RXB. Spent Fuel Storage Facility 2 Partially Conforms The design of the new and spent fuel storage facility 3.2 Design Basis complies with Regulatory Position C.8, Makeup 9.1 Water by the large inventory of water within the 9.2 Seismic Category I structures forming the ultimate 3.5.2 heat sink (UHS) and by the separate Quality Group C, Seismic Category I makeup line. For Regulatory Posi-tion C.9, Pool Cooling, the UHS pool structures con-taining the inventory of makeup water credited for spent fuel cooling during accident conditions meet Seismic Category I requirements, but as structures, they are not designed to Quality Group C require-ments. In addition, the reactor building ventilation system is not credited with the capability to vent steam or moisture to the atmosphere to protect safety-related components from high temperatures and moisture levels because such protection is not required for the design. Reactor Coolant Pump Flywheel 1 Not Applicable This guidance is applicable only to PWR designs that Not Applicable Integrity rely on reactor coolant pumps. The NuScale design uses passive natural circulation of the primary cool-ant, eliminating the need for reactor coolant pumps. Comprehensive Vibration Assess- 3 Conforms The first operational NPM is classified as a prototype 3.9 Conformance with Regulatory Criteria ment Program for Reactor Inter- in accordance with RG 1.20. Thus, the portions of 5.4 nals During Preoperational and this RG which apply to prototype reactors are appli- 14.2 Initial Startup Testing cable to the first operational NPM. After the first NPM is qualified as a valid prototype, subsequent NPMs are classified as non-prototype category I and the non-prototype portions of the RG apply. The identification of departures from RG 1.20 is the responsibility of the COL applicant or licensee.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Measuring, Evaluating, and 2 Partially Conforms Site-specific, programmatic and operational aspects 11.5 Reporting Radioactive Material in are the responsibility of the COL applicant or Liquid and Gaseous Effluents and licensee. Solid Waste Periodic Testing of Protection 0 Conforms None. 7.2 System Actuation Functions Meteorological Monitoring Pro- 1 Not Applicable This guidance is the responsibility of the COL appli- Not Applicable grams for Nuclear Power Plant cant or licensee. Assumptions Used for Evaluating 0 Not Applicable Site-specific guidance is the responsibility of the Not Applicable the Potential Radiological Conse- COL applicant or licensee. quences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure Assumptions Used for Evaluating 0 Not Applicable This RG pertains to TID14844 source terms used by Not Applicable the Potential Radiological Conse- licensees of existing reactors that are not authorized quences of a Fuel Handling Acci- to use the alternative source term under 10 CFR dent in the Fuel Handling and 50.67. RG 1.183 is specified to be used in lieu of RG Storage Facility for Boiling and 1.25 for new reactors (and existing reactors autho-Pressurized Water Reactors rized to use the alternative source term under 10 CFR 50.67). Quality Group Classifications and 4 Conforms The quality group classification from RG 1.26 appli- 3.2 Standards for Water-, Steam-, and cable to a specific component is described through- 5.2 Radioactive-Waste-Containing out the FSAR. 5.4 Components of Nuclear Power Plants 6.2 9.1 Conformance with Regulatory Criteria 9.2 9.3 10.3 10.4 Ultimate Heat Sink for Nuclear 3 Not Applicable RG does not apply to plants that use a passive cool- Not Applicable Power Plants ing system to transfer heat to the ultimate heat sink. The NuScale design uses a passive cooling system.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Quality Assurance Program Crite- 4 Conforms None. 3.13 ria (Design and Construction) 4.5 5.2 5.3 5.4 6.1 7.0 7.2 14.3 17.1 17.5 Seismic Design Classification for 5 Partially Conforms Each SSC described in Staff Regulatory Guidance 3.2 Nuclear Power Plants C.1.a through C.1.h is designated as Seismic Cate-gory I. SSC that meet Staff Regulatory Guidance C.1.i are designated Seismic Category II rather than Seis-mic Category I. The seismic classification from RG 1.29 applicable to a specific component is described throughout the FSAR. Safety Guide 30 - Quality Assur- - Not Applicable This RG endorses IEEE Std. 336-1971 for the installa- Not Applicable ance Requirements for the Instal- tion, inspection, and testing of instrumentation and lation, Inspection, and Testing of electric equipment. The NuScale design is based on Instrumentation and Electric NQA-1-2008 and the NQA-1a-2009 addenda, as Equipment endorsed in RG 1.28, Rev. 4. NQA-1-2008 and NQA-1a-2009 (Subpart 2.4) reference IEEE Std. 336-Conformance with Regulatory Criteria 1985 (as opposed to IEEE Std. 336-1971). The sub-stantive content and intent of RG 1.30 is contained in Subpart 2.4 of NQA-1-2008 and NQA-1a-2009 and IEEE Std. 336-1985, which is applicable to the NuS-cale design per NQA-2008 and NQA-1a-2009. Control of Ferrite Content in 4 Conforms None. 4.5 Stainless Steel Weld Metal 5.2 6.1 Criteria for Power Systems for 3 Partially Conforms RG 1.32 is not applicable to the offsite and onsite AC 8.2 Nuclear Power Plants power systems. The EDSS conforms to RG 1.32 to the 8.3 extent described in Section 8.3.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Quality Assurance Program 3 Not Applicable This guidance is the responsibility of the COL appli- Not Applicable Requirements (Operation) cant or licensee. Control of Electroslag Weld Prop- 1 Conforms None. 5.3 erties Inservice Inspection of 3 Not Applicable The NuScale design uses a steel containment vessel Not Applicable Ungrouted Tendons in Pre- (i.e., does not use concrete in its design). stressed Concrete Containments 1 Determining Prestressing Forces - Not Applicable The NuScale design uses a steel containment vessel Not Applicable for Inspection of Prestressed Con- (i.e., does not use concrete in its design). crete Containments Nonmetallic Thermal Insulation - Not Applicable The NuScale design does not use nonmetallic ther- Not Applicable for Austenitic Stainless Steel mal insulation on RCPB or CNV components. Qualification of Continuous Duty 1 Not Applicable The NuScale design does not use continuous duty Not Applicable Safety-Related Motors for Nuclear Class 1E motors. Power Plants Preoperational Testing of Redun- - Partially Conforms Portions of this guide are applicable to preopera- 8.3 dant Onsite Electric Power Sys- tional testing of divisional EDSS load groups. It does 14.2 tems to Verify Proper Load Group not apply to NuScale AC power systems or the EDNS. Assignments Control of Stainless Steel Weld 1 Conforms None. 5.2 Cladding of Low-Alloy Steel Com- 5.3 ponents 6.1 Control of the Processing and Use 1 Partially Conforms This RG is applicable except for its specification of 4.5 of Stainless Steel applying RG 1.37 for cleaning and flushing of fin- 5.2 ished surfaces. RG 1.37 has been withdrawn by the 5.3 Conformance with Regulatory Criteria NRC. 6.1

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Guidance on Monitoring and 1 Conforms The design satisfies RG 1.45 guidance by using two 3.6 Responding to Reactor Coolant systems to detect leakage into the containment: 5.2 System Leakage containment pressure monitoring and leakage col- 6.2 lection. Both leakage detection methods satisfy Reg-9.3 ulatory Positions C.2.1 and C.2.2 in RG 1.45: a) leakage to the primary reactor containment from 11.5 unidentified sources can be detected, monitored, 14.2 and quantified for rates 0.05 gpm; and; b) 14.3 response time (not including transport delay time) is less than one hour for a leakage rate greater than one gpm. Regulatory Position C.2.4 is satisfied because the containment pressure method is capa-ble of performing its function following a seismic event that does not require plant shutdown (i.e., vacuum pump remains functional). C.2.5 is satisfied because both methods permit calibration and test-ing during plant operation. Finally, radiation detec-tors in the CES condenser vent line provide an early indication of RCS leakage consistent with Regulatory Position C.2.3. All leakage is treated as unidentified because of the limited capability to identify or quan-tify RCS leakage. Bypassed and Inoperable Status 1 Conforms None. 7.2 Indication for Nuclear Power Plant Safety Systems Control of Preheat Temperature 1 Conforms None. 5.2 for Welding of Low-Alloy Steel Conformance with Regulatory Criteria 6.1

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Design, Inspection, and Testing 4 Not Applicable This guidance addresses engineered safety feature Not Applicable Criteria for Air Filtration and (ESF) filter and atmosphere cleanup systems Adsorption Units of Post-Acci- designed for fission product removal in a post-dent Engineered-Safety-Feature design basis accident environment. The NuScale Atmosphere Cleanup Systems in design does not rely on ESF filter and atmosphere Light-Water-Cooled Nuclear cleanup systems to mitigate the consequences of a Power Plants design basis accident. Nonsafety-related normal ventilation systems provide atmosphere cleanup capability, as necessary, that meets the design, test-ing, and maintenance guidelines in RG 1.140. These systems provide appropriate containment, confine-ment, and filtering to limit releases of airborne radio-activity to the environment during normal and postulated accident conditions. However, these sys-tems are not required following an accident at a NuScale Power Plant, and accordingly receive no credit in the determination of the radiological con-sequences of an accident. Application of the Single-Failure 2 Conforms None. 7.1 Criterion to Safety Systems 8.3 15.1 15.2 15.5 Service Level I, II, and III Protective 2 Partially Conforms Applicable except for portions of this RG that govern 11.2 Coatings Applied to Nuclear operational aspects (e.g., maintenance of safety-Power Plants related coatings) that are the responsibility of the Conformance with Regulatory Criteria COL applicant or licensee.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Design Limits and Loading Com- 2 Partially Conforms Applicable except for reference to 3.8 binations for Metal Primary Reac- 10 CFR 50.34(f)(3)(v), since per 10 CFR 50.34(f) and 6.2 tor Containment System 10 CFR 52.47(a)(8), a design certification applicant Components does not have to show compliance with 10 CFR 50.34(f)(3)(v). Use of typical reactor pressure vessel load combinations for Class 1 vessels is more applicable to the containment vessel than using the load combinations specified in RG 1.57 because of the increased quality of the fabrication, inspection, and testing required by American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel ASME Code, Section III, Subsection NB for a Class 1 vessel. Additionally, the use of load combinations for Class 1 vessels is more applicable to the containment vessel than using the load com-binations specified in RG 1.57 because of the quality of the fabrication, inspection, and testing required by ASME Code, Section III, Subsection NB for a Class 1 vessel. Design Basis Floods for Nuclear 2 Not Applicable The NuScale design assumes the NPP is located Not Applicable Power Plants above the probable maximum flood height (includ-ing wind-induced wave run-up). Design Response Spectra for Seis- 2 Not Applicable The Certified Seismic Design Response Spectra Not Applicable mic Design of Nuclear Power (CSDRS) was not developed using RG 1.60. However, Plants it is demonstrated that the design envelops the RG 1.60 spectra anchored to 0.1g. Damping Values for Seismic 1 Conforms In accordance with the guidance of RG 1.61, an alter- 3.7 Conformance with Regulatory Criteria Design of Nuclear Power Plants native damping value for the NPM substructure was 3.8 determined. 3.12 Manual Initiation of Protective 1 Conforms None. 7.1 Actions 7.2 Electric Penetration Assemblies in 3 Partially Conforms The portion of the RG 1.63 guidance that endorses 3.8.2 Containment Structures for IEEE-317-1983 is applicable. IEEE 741-1997 is used 3.11 Nuclear Power Plants for external circuit protection of electrical penetra- 8.1 tion assemblies instead of IEEE 741-1986 as 8.3 endorsed by RG 1.63. The 1997 version, including the additional design enhancements, is consistent 14.3 with RG 1.63.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Materials and Inspections for 1 Partially Conforms Inservice inspection is the responsibility of the COL 3.13 Reactor Vessel Closure Studs applicant. The reactor pressure vessel (RPV) bolting 5.3 material is described in Table 5.2-4. The material is not subject to the concerns addressed by RG 1.65 Positions 1(a)(i) and 2(b). Therefore, these Positions do not apply to the RPV bolting material. Initial Test Programs for Water- 4 Partially Conforms This guidance is applicable except for aspects that 4.4 Cooled Nuclear Power Plants are BWR-specific or address specific PWR SSC design 5.4 features not in the NuScale design. Site-specific pro-8.2 gram implementation activities are the responsibil-ity of the COL applicant or licensee that references 8.3 the NuScale certified design. 9.3.2 10.4 14.2 14.3 1 Initial Test Program of Conden- 2 Partially Conforms This RG is applicable except for aspects that are 14.2 sate and Feedwater Systems for BWR-specific or address specific PWR design fea-Light-Water Reactors tures not in the NuScale design. 2 Initial Startup Test Program to 2 Partially Conforms This guidance is applicable except for site-specific 14.2 Demonstrate Remote Shutdown aspects including test performance, test report Capability for Water-Cooled preparation, and records retention, which are the Nuclear Power Plants responsibility of the COL applicant or licensee. 3 Preoperational Testing of Instru- 1 Partially Conforms This guidance is applicable except for site-specific 14.2 ment and Control Air Systems aspects, including test performance and records retention, which are the responsibility of the COL Conformance with Regulatory Criteria applicant or licensee. Concrete Radiation Shields and 1 Partially Conforms This guidance is applicable to the design of concrete 3.8 Generic Shield Testing for Nuclear radiation shields. Site-specific aspects of this guid- 12.3 Power Plants ance, including development and implementation 14.2 of a radiation shield test program, are the responsi-bility of the COL applicant or licensee. Standard Format and Content of 3 Not Applicable RG 1.206 and NuScale Design Specific Review Stan- Not Applicable Safety Analysis Reports for dards (DSRS) are used. Nuclear Power Plants (LWR Edi-tion)

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Welder Qualification for Areas of 1 Partially Conforms This guidance is applicable except for site-specific 4.5 Limited Accessibility aspects, including specification of standards for 5.2 weld fabrication and repair that are performed 6.1 during construction, installation, and operation of a nuclear facility, which are the responsibility of the COL applicant or licensee. Spray Pond Piping Made from 2 Not Applicable The design does not use fiberglass piping in spray Not Applicable Fiberglass-Reinforced Thermoset- pond applications (or for the UHS design). ting Resin Qualification Tests for Safety- 1 Partially Conforms The guidance is applicable except for portions that 3.11 Related Actuators in Nuclear apply to high temperature gas-cooled reactor Power Plants designs. Criteria for Independence of Elec- 3 Conforms None. 7.1 trical Safety Systems 7.2 8.3 9.5 14.3 Design-Basis Tornado and Tor- 1 Conforms Region 1 (bounding) characteristics are used as 2.3 nado Missiles for Nuclear Power design parameters. The COL applicant or licensee is 3.3 Plants responsible for confirming the characteristics. 3.5 3.8 5.0 Conformance with Regulatory Criteria

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Assumptions Used for Evaluating - Partially Conforms Portions of this RG pertain to assumptions for radio- 15.0.3 a Control Rod Ejection Accident logical consequence analysis. Per SRP Section 15.0.3, 15.4 for Pressurized Water Reactors Section I, Areas of Review, Item 10 under subhead-ing Review Interfaces, for the review of design certi-fication applications, SRP Section 15.0.3 supersedes the radiological analyses, assumptions, acceptance criteria, and methodologies identified in SRP Section 15.4.8, which references guidance in RG 1.77. The NRC has identified this RG as out of date and in need of revision. The fuel and cladding failure crite-ria are superseded by the criteria provided SRP (NUREG-0800) 4.2, Appendix B. The radiological cri-teria are superseded by the criteria in RG 1.183. How-ever, the general approach and intent of RG 1.77 still apply and are used in Section 15.4.8 analyses. Evaluating the Habitability of a 1 Partially Conforms Aspects of this RG related to control room habitabil- 3.2 Nuclear Power Plant Control ity design within the scope of the NuScale design are 6.4 Room During a Postulated Haz- applicable. Other aspects of this guidance require 9.4 ardous Chemical Release site-specific information (e.g., amount and location of toxic chemicals relative to the control room, and site-specific atmospheric dispersion factors) or spec-ify operational, programmatic emergency planning activities. These aspects are the responsibility of the COL applicant or licensee. Preoperational Testing of Emer- 2 Partially Conforms The intent of this RG is applicable to the NuScale 6.3 gency Core Cooling Systems for design, but the literal language refers to SSC design 14.2 Conformance with Regulatory Criteria Pressurized Water Reactors features not in the NuScale design. For example, the ECCS design does not use high pressure or low pres-sure safety injection pumps as described in this guidance. Rather, the ECCS design provides core decay heat removal by steam condensation and nat-ural reactor coolant recirculation. Nevertheless, pre-operational testing will be performed on the ECCS in a manner that satisfies the intent of this guidance. Much of this RG prescribes preoperational test implementation activities that are the responsibility of the COL applicant.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 1 Initial Test Program of Emergency - Not Applicable RG 1.79.1 is applicable to BWRs only. Not Applicable Core Cooling Systems for New Boiling-Water Reactors Shared Emergency and Shut- 1 Not Applicable RG 1.81 is not relevant to the AC power systems. As Not Applicable down Electric Systems for Multi- described in Section 8.3, the EDSS conforms to por-Unit Nuclear Power Plants tions of RG 1.81. Water Sources for Long-Term 4 Partially Conforms The NuScale design complies with the intent of RG 6.3 Recirculation Cooling Following a 1.82 regulatory positions that address the design Loss-of-Coolant Accident criteria, performance standards, and analysis methods related to water sources for long-term cooling. However, the NuScale design differs from the system designs the guidance addresses. The NuScale design complies with the guidance with respect to debris generation, debris transport, coating debris, latent debris, downstream, and chemical effects. The NuScale design is passive and does not include pumps, sumps, suction strainers, debris interceptor, or trash racks, and the design minimizes or negates the potential effect of non-condensables on coolant flow to the core. The NuScale design does not require operator action to mitigate debris accumulation. The NuScale design does not comply with regulatory position C1.1 with the exception that NuScale does comply with the intent of the following regulatory positions: Conformance with Regulatory Criteria

• Position C1.1.1.9 (assessment of the possibility of downstream clogging).
• Position C1.1.1.10 (buildup of debris and chemical reaction products downstream).
• Position C.1.1.2 (minimization of debris source term, cleanness programs, monitoring/sampling for latent debris, insulation selection, restriction on coatings and cladding of carbon steel).

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Positions C1.1.3 and C1.1.4 are not applicable because the NuScale design does not rely on operator action to mitigate the consequences of debris accumulation and does not include active devices or systems to prevent debris accumulation. The NuScale design does not comply with regulatory position C.1.2 (alternative water sources for inoperable strainers). The NuScale design complies with the intent of regulatory position C.1.3 (evaluation of long term recirculation capability as applicable to the design) with the exception of the following:

• Position C.1.3.1 (net positive suction head)
• Portions of position C.1.3.2 that are not consistent with the NuScale design The NuScale design does not comply with regulatory positions C1.3.7 (upstream effects) or C.1.3.9 (strainer structural integrity).

The NuScale design does not comply with regulatory position C1.3.12 (prototypical head loss testing). The NuScale design does not comply with regulatory position C.2 with the exception that the intent of chemical reaction effects (position 2.2) is met. Conformance with Regulatory Criteria The NuScale design does not comply with regulatory position C.3. Design, Fabrication, and Materials 36 Conforms None. 3.12 Code Case Acceptability, ASME 3.13 Section III 4.5 5.2 Termination of Operating - Not Applicable This RG governs the process for terminating nuclear Not Applicable Licenses for Nuclear Reactors reactor operating licenses.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Guidance for Construction of 1 Not Applicable This RG applies to elevated-temperature reactors Not Applicable Class 1 Components in Elevated- such as high-temperature gas-cooled reactors, liq-Temperature Reactors (Supple- uid-metal fast-breeder reactors, and gas-cooled fast-ment to ASME Section III Code breeder reactors. Cases 1592, 1593, 1594, 1595, and 1596) Environmental Qualification of 1 Partially Conforms This RG is applicable except for: (1) aspects that are 3.8.2 Certain Electric Equipment BWR-specific or related to SSC that are not relevant 3.11 Important to Safety for Nuclear to the NuScale design (e.g., ice condenser contain- Appendix 3C Power Plants ment, containment spray system, etc.); and (2) refer-ence to RG 1.4 for source term, since the source term provisions of RG 1.4 are superseded by RG 1.183 for new reactors. Inservice Inspection of Pre- 2 Not Applicable This RG is applicable to LWR designs that incorpo- Not Applicable stressed Concrete Containment rate a pre-stressed concrete containment structure Structures with Grouted Tendons with grouted tendons. The NuScale containment vessel is steel (i.e., does not use concrete or grouted tendons in its design). Evaluations of Explosions Postu- 2 Not Applicable This guidance governs the performance of site-spe- Not Applicable lated to Occur on Transportation cific evaluations and is the responsibility of the COL Routes Near Nuclear Power Plants applicant or licensee. Combining Modal Responses and 3 Conforms None. 3.7 Spatial Components in Seismic 3.8 Response Analysis 3.9 3.10 Conformance with Regulatory Criteria 3.12 Availability of Electric Power 1 Not Applicable This RG is not identified as an applicable RG in DSRS Not Applicable Sources Section 8.1. Design of Main Steam Isolation 1 Not Applicable This RG is applicable only to BWR designs. Not Applicable Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Criteria for Accident Monitoring 4 Partially Conforms The NuScale design satisfies power supply require- 3.11 Instrumentation for Nuclear ments in Section 6.6 of IEEE Std 497-2002 for Type B Appendix 3C Power Plants and C variables with highly reliable power rather 5.4 than with Class 1E. The portions of RG 1.97 dealing 7.1 with 10 CFR 50.34(f)(2)(xix) are addressed in 7.2 Section 19.2.3.3.8. 8.3 11.5 12.3 14.3 19.2 Assumptions Used for Evaluating 0 Not Applicable This RG is applicable only to BWR designs. Not Applicable the Potential Radiological Conse-quences of a Radioactive Offgas System Failure in a Boiling Water Reactor Radiation Embrittlement of Reac- 2 Conforms None. 5.3 tor Vessel Materials 0 Seismic Qualification of Electrical 3 Partially Conforms This RG is applicable except for aspects related to: (1) 3.9 and Active Mechanical Equip- when site-specific spectra exceed the certified 3.10 ment and Functional Qualifica- design spectra (e.g., Position C1.2.1.g); and (2) quali-5.2 tion of Active Mechanical fication of new and replacement equipment in older Equipment for Nuclear Power unresolved safety issue A46 plants (e.g., Position 14.3 Plants C.1.2.2.j). Not applicable to electrical equipment. Appendix 3C Site-specific guidance is the responsibility of the COL applicant. RG 1.100 endorses Conformance with Regulatory Criteria ASME QME-1 2007. NuScale complies with the non-mandatory Appendix QR-B with the following exceptions: QR-B5200, Identification and Specification of Qualifi-cation Requirements, (g) material activation energy. QR-B5300 Selection of Qualification Methods for determination and recording of shelf life of nonmetallics. QR-B5500 Documentation, (h) shelf life preservation requirements. Appendix 3C describes the exceptions cited.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 1 Emergency Response Planning 5 Not Applicable This RG is the responsibility of the COL applicant Not Applicable and Preparedness for Nuclear proposing to site a power plant that meets the defi-Power Reactors nition of co-located. 2 Flood Protection for Nuclear 1 Applicable The design assumes the NPP is located above the 2.4 Power Plants probable maximum flood height (including wind 3.4 induced wave run-up). 3.8 5 Setpoints for Safety-Related 3 Partially Conforms Chapter 15 analyses use the safety-related setpoints 7.2 Instrumentation described in Chapter 7. This RG endorses ISA- 15.1 67.04.01-1994, however, the NuScale Instrument 15.2 Setpoint Methodology Technical Report (TR-0616-49121) applies the guidance contained in ISA- 15.4 67.04.01-2006. A key difference is that the 1994 ver- 15.5 sion of ISA-67.04.01 uses an allowable value to 15.6 determine instrument channel operability during surveillance testing and calibration. The 2006 ver-sion of ISA-67.04.01 provides updated guidance for evaluating instrument channel operability based on the comparison of the as-found to the as-left value from the previous instrument calibration for the instrument setpoint. 6 Thermal Overload Protection for 2 Not Applicable This RG governs the application of thermal overload Not Applicable Electric Motors on Motor-Oper- protection devices to ensure that safety-related ated Valves motor-operated valves perform their safety func-tion. The NuScale design does not use safety-related motor-operated valves. 7 Qualification for Cement Grout- 2 Not Applicable This RG is applicable only to LWR designs that use a Not Applicable Conformance with Regulatory Criteria ing for Prestressing Tendons in prestressed concrete containment structure. The Containment Structures containment vessel is a steel containment (i.e., does not use concrete or pre-stressed tendons in its design). 9 Calculation of Annual Doses to 1 Partially Conforms This RG is applicable except for specification of site- 11.2 Man from Routine Releases of specific information (e.g., meteorological data). Site- 11.3 Reactor Effluents for the Purpose specific information is the responsibility of the COL of Evaluating Compliance with 10 applicant. CFR Part 50, Appendix I

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 0 Cost-Benefit Analysis for Rad- 1 Partially Conforms This RG is applicable except for aspects related to 11.2 waste Systems for Light-Water- performance of a site-specific cost-benefit analysis. 11.3 Cooled Nuclear Power Reactors Site-specific information is the responsibility of the COL applicant or licensee. 1 Methods for Estimating Atmo- 1 Partially Conforms This RG is applicable except for specification of site- 2.3 spheric Transport and Dispersion specific dispersion data. Site-specific information is 3.3 of Gaseous Effluents in Routine the responsibility of the COL applicant or licensee. Releases from Light-Water-Cooled Reactors 2 Calculation of Releases of Radio- 1 Partially Conforms This RG is applicable except for specification of site- 2.3 active Materials in Gaseous and specific information (e.g., meteorological data). Site- 11.2 Liquid Effluents from Light- specific information is the responsibility of the COL 11.3 Water-Cooled Nuclear Power applicant or licensee. Reactors 3 Estimating Aquatic Dispersion of 1 Not Applicable This RG provides guidance for analyzing the aquatic Not Applicable Effluents from Accidental and dispersion of radioactive liquid effluents from com-Routine Reactor Releases for the ponent failures, in accordance with BTP 11-6. Purpose of Implementing Appen- Because the NuScale facility provides an approved dix I design mitigative feature (metal-lined concrete dike around the PSCS storage tank), such an analysis is not required. 4 Guidance to Operators at the 3 Partially Conforms Site-specific guidance is the responsibility of the 18.5 Controls and to Senior Operators COL applicant or licensee. Consistent with the dis-in the Control Room of a Nuclear cussion in RG 1.114, Section B.1, the ability of the Power Unit COL applicant to meet this guidance is facilitated by the control room design and layout (including the Conformance with Regulatory Criteria designated surveillance area described in Position C.1.3). Portions of this guidance that implement operator staffing requirements of 10 CFR 50.54(m)(2)(i) and (iii) (e.g., Position C.1.5) are not applicable to COL applicants. 5 Protection Against Turbine Mis- 2 Conforms None. 3.5.3 siles 7 Protection Against Extreme Wind 2 Conforms Confirmation that nearby structures exposed to 3.5 Events and Missiles for Nuclear extreme wind loads will not adversely affect the 9.1 Power Plants Reactor Building or the Seismic Category I portion of the Control Building is the responsibility of the COL applicant or licensee.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 8 Periodic Testing of Electric Power 3 Partially Conforms Site-specific guidance is the responsibility of the 7.2 and Protection Systems COL applicant or licensee. 8.3 14.2 1 Bases for Plugging Degraded August 1976 Conforms None. 5.4 PWR Steam Generator Tubes 2 Development of Floor Design 1 Partially Conforms Site-specific guidance is the responsibility of the 3.7 Response Spectra for Seismic COL applicant or licensee. 3.12 Design of Floor-Supported Equip-ment or Components 4 Service Limits and Loading Com- 3 Conforms None. 3.9 binations for Class 1 Linear-Type Supports 5 Physical Models for Design and 2 Not Applicable The NuScale design does not require hydraulic struc- Not Applicable Operation of Hydraulic Struc- tures. tures and Systems for Nuclear Power Plants 6 An Acceptable Model and Related 2 Conforms None. 4.2 Statistical Methods for the Analy-sis of Fuel Densification 7 Inspection of Water-Control 1 Not Applicable This guidance governs the development of an inser- Not Applicable Structures Associated with vice inspection and surveillance program for dams, Nuclear Power Plants slopes, canals, and other water-control structures associated with emergency cooling water systems or flood protection of nuclear power plants. Water control structures and associated inservice inspec-Conformance with Regulatory Criteria tion and surveillance programs are site-specific details. Site-specific guidance is the responsibility of the COL applicant or licensee. 8 Installation Design and Installa- 2 Partially Conforms The EDSS uses VRLA batteries; thus 8.3 tion of Vented Lead-Acid Storage IEEE Std 1187-2013 is applied. 14.2 Batteries for Nuclear Power Plants 9 Maintenance, Testing, and 3 Partially Conforms The EDSS uses VRLA batteries. NuScale applies 8.3 Replacement of Vented Lead- IEEE Std 1188-2005 with the 2014 amendment. Acid Storage Batteries for Nuclear Power Plants

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 0 Service Limits and Loading Com- 3 Conforms None. 3.9 binations for Class 1 Plate-and-Shell-Type Supports 2 Site Investigations for Founda- 2 Not Applicable This RG governs site investigations performed as Not Applicable tions of Nuclear Power Plants part of site selection. Site-specific guidance is the responsibility of the COL applicant or licensee. 3 Loose-Part Detection Program for 1 Not Applicable The low fluid velocities resulting from natural circu- Not Applicable the Primary System of Light- lation flow combined with a design that has only Water-Cooled Reactors small lines entering the RPV minimizes the potential for loose parts entering or being generated in the RPV. Additional justification for this information is in Section 4.4 of the FSAR. 4 Medical Evaluation of Licensed 3 Not Applicable This RG governs site-specific operational program Not Applicable Personnel at Nuclear Power activities. Compliance with site-specific guidance is Plants the responsibility of the COL applicant or licensee. 6 Design Limits, Loading Combina- 3 Not Applicable This guidance is applicable only to LWR designs that Not Applicable tions, Materials, Construction, use concrete containments. The NuScale design and Testing of Concrete Contain- uses a steel containment vessel. ments 7 Fuel-Oil Systems for Standby Die- 2 Not Applicable The design does not rely on or include safety-related Not Applicable sel Generators emergency diesel generators. 8 Laboratory Investigations of Soils 2 Not Applicable This guidance is related to site-specific laboratory Not Applicable and Rocks for Engineering Analy- investigation activities. Site-specific guidance is the sis and Design of Nuclear Power responsibility of the COL applicant or licensee. Plants 0 Design, Inspection, and Testing 2 Partially Conforms Design-related aspects of this guidance are applica- 3.2 Conformance with Regulatory Criteria Criteria for Air Filtration and ble. Aspects related to construction, testing, and 9.4 Adsorption Units of Normal repairs are the responsibility of the COL applicant or 11.3 Atmosphere Cleanup Systems in licensee. 12.3 Light-Water-Cooled Nuclear Power Plants 14.2 1 Containment Isolation Provisions 1 Conforms The NuScale design conforms to the requirements of 6.2 for Fluid Systems RG 1.141 through adherence to ANS N271-1976. Note: the provisions of ANSI/ANS 56.2-1984, Section 3.6.5 are applied to penetrations with both CIVs outside containment that serve non-ESF process systems.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 2 Safety-Related Concrete Struc- 2 Partially Conforms The intent of this guidance is applicable but the lan- 3.8 tures for Nuclear Power Plants guage endorses ACI 349-1997 with exceptions. The (Other than Reactor Vessels and 2006 version of the ACI 349 standard has been used. Containments) Aspects of Regulatory Positions C.1 and C.14 related to concrete structures within containment are not applicable to the design. The containment vessel is a steel component, and does not use interior concrete structures. 3 Design Guidance for Radioactive 2 Partially Conforms The aspects of this RG related to steam generator 3.2 Waste Management Systems, blowdown systems are not applicable to the design. 3.3 Structures, and Components Radioactive waste management system design crite-3.5 Installed in Light-Water-Cooled ria specified in this RG are applicable. Construction, Nuclear Power Plants installation, and testing criteria are the responsibility 3.7 of the COL applicant or licensee. 9.2.6 11.2 11.3 11.4 14.3 5 Atmospheric Dispersion Models 1 Not Applicable This RG does not include modeling of building wake Not Applicable for Potential Accident Conse- effects. For the short distances that may be used for quence Assessments at Nuclear EAB and LPZ, Regulatory Guide 1.194 is used to Power Plants determine representative atmospheric dispersion factors. 7 Inservice Inspection Code Case 17 Partially Conforms Performance of inservice inspections per the 5.2 Acceptability, ASME Section XI, American Society of Mechanical Engineers Boiler 6.6 Conformance with Regulatory Criteria Division 1 and Pressure Vessel Code is the responsibility of the COL applicant or licensee. 9 Nuclear Power Plant Simulation 4 Not Applicable Simulation facilities and conduct of licensed opera- Not Applicable Facilities for Use in Operator tor training and qualification are the responsibility of Training, License Examinations, the COL applicant or licensee. and Applicant Experience Requirements 1 Instrument Sensing Lines 1 Partially Conforms This RG governs design and installation of safety- 7.2 related instrument sensing lines in nuclear power plants. The aspects of this RG regarding installation criteria are the responsibility of the COL applicant or licensee.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 2 Criteria for Use of Computers in 3 Partially Conforms The NuScale I&C development lifecycle differs from 3.11 Safety Systems of Nuclear Power the conceptual waterfall lifecycle in RG 1.152. The 7.1 Plants applicable tasks from the RG lifecycle model will be 7.2 mapped to the I&C development lifecycle. Compli-ance with Clause 5.5 of IEEE 7-4.3.2-2003 is condi- 14.3 tioned by the choice of field programmable gate array technology, which makes some tests not appli-cable (e.g., calculation tests, watchdog timer tests). 3 Criteria for Safety Systems 1 Conforms Applicable to EDSS. 3.11 8.3 5 Station Blackout 1 Partially Conforms The design conforms to the aspects of the RG as it 5.4 pertains to passive plant designs. 6.2 8.3 8.4 9.3 9.5 10.3 14.2 6 Qualification of Connection 1 Conforms None. 3.11 Assemblies for Nuclear Power Plants 7 Best-Estimate Calculations of - Not Applicable Best estimate calculations are not used. Not Applicable Emergency Core Cooling System Performance Conformance with Regulatory Criteria 8 Qualification of Safety-Related - Conforms The DC system batteries are non-Class 1E. 3.11 Lead Storage Batteries for Nuclear Power Plants 9 Assuring the Availability of Funds 2 Not Applicable Decommissioning funding activities are the respon- Not Applicable for Decommissioning Nuclear sibility of the COL applicant. Reactors 0 Monitoring the Effectiveness of 3 Not Applicable Monitoring the effectiveness of maintenance activi- Not Applicable Maintenance at Nuclear Power ties is the responsibility of the COL applicant. Plants

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 1 Evaluation of Reactor Pressure - Not Applicable The Charpy upper-shelf energy of the NuScale reac- Not Applicable Vessels with Charpy Upper-Shelf tor vessel materials will exceed the 50 ft-lb energy Energy Less Than 50 Ft-Lb value (throughout the life of the vessel with signifi-cant margin) below which this guidance would apply. However, in the unlikely event the reactor vessel material surveillance program implemented during reactor operations indicates that this is not the case, the requirements of 10 CFR 50, Appendix G and the provisions of RG 1.161 would be the respon-sibility of the COL applicant or licensee (see discus-sion of RG 1.162). 2 Format and Content of Report for - Not Applicable If thermal annealing becomes necessary, the Not Applicable Thermal Annealing of Reactor requirements of 10 CFR 50.66 and the provisions of Pressure Vessels RG 1.162 would be the responsibility of the COL applicant or licensee. 3 Performance-Based Containment - Not Applicable The design of containment penetrations supports Not Applicable Leak-Test Program performance of local leak rate tests (Type B and Type C tests) in accordance with the guidance provided in ANSI/ANS 56.8, Regulatory Guide 1.163, and NEI 94-

01. The NuScale system design, in conformance with 10 CFR 50.54(o), accommodates the 10 CFR 50, Appendix J, test method frequencies of Option A or Option B. This RG is the responsibility of a COL appli-cant or licensee that seeks to implement Option B.

6 Pre-Earthquake Planning and - Not Applicable This RG governs programmatic activities (earth- Not Applicable Immediate Nuclear Power Plant quake planning and post-earthquake actions) that Operator Post Earthquake Actions are the responsibility of the COL applicant or Conformance with Regulatory Criteria licensee. 7 Restart of a Nuclear Power Plant - Not Applicable This RG governs post-earthquake inspections and Not Applicable Shut Down by a Seismic Event tests that are the responsibility of the COL applicant or licensee.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 8 Verification, Validation, Reviews, 2 Partially Conforms The NuScale design applies RG 1.152, Revision 3 and 7.2 and Audits for Digital Computer IEEE Std. 7-4.3.2-2003 that it endorses. For RG 1.168, 14.3 Software Used in Safety Systems the requirements of IEEE 1012-2004 are tailored to of Nuclear Power Plants the NuScale I&C development lifecycle, which is dif-ferent from the conceptual waterfall lifecycle in IEEE 1012-2004. The applicable tasks from IEEE 1012-2004 to the I&C development are mapped. Some administrative mandatory require-ments in the standard conflict with established Engi-neering or QA documentation requirements. The requirements of IEEE 1028-2008 are tailored to the NuScale I&C development lifecycle. 9 Configuration Management Plans 1 Partially Conforms For this RG, the requirements of IEEE 828-2005 are 7.2 for Digital Computer Software tailored to the NuScale I&C development lifecycle, 14.3 Used in Safety Systems of Nuclear which is different from the conceptual waterfall life-Power Plants cycle in RG 1.152. The applicable tasks from IEEE 828-2005 are mapped to the NuScale I&C devel-opment lifecycle. 0 Test Documentation for Digital 1 Partially Conforms Requirements of IEEE 829-2008 are tailored to the 7.2 Computer Software Used in NuScale I&C development lifecycle, which is differ- 14.3 Safety Systems of Nuclear Power ent from the conceptual waterfall lifecycle in Plants RG 1.152. The applicable tasks from IEEE 829-2008 are mapped to the NuScale I&C development lifecy-cle. NuScale takes exception to some administrative mandatory requirements in the standard that con-flict with established Engineering or quality docu-mentation requirements. Conformance with Regulatory Criteria 1 Software Unit Testing for Digital 1 Partially Conforms NuScale takes exception to some administrative 7.2 Computer Software Used in mandatory requirements in IEEE 1008-1987 that 14.3 Safety Systems of Nuclear Power conflict with established Engineering or quality doc-Plants umentation requirements. 2 Software Requirements Specifica- 1 Partially Conforms NuScale takes exception to some administrative 7.2 tions for Digital Computer Soft- mandatory requirements in IEEE 830-1998 standard 14.3 ware Used in Safety Systems of that conflict with established Engineering or quality Nuclear Power Plants documentation requirements.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 3 Developing Software Life Cycle 1 Partially Conforms Requirements of IEEE 1074-2006 are tailored to the 7.2 Processes for Digital Computer NuScale I&C development lifecycle, which differs 14.3 Software Used in Safety Systems from the conceptual waterfall lifecycle in RG 1.152. of Nuclear Power Plants Applicable tasks from IEEE 1074-2006 are mapped to the NuScale I&C development lifecycle. NuScale takes exception to some administrative mandatory requirements in the standard that conflict with established Engineering or quality documentation requirements. 4 An Approach for Using Probabilis- 2 Not Applicable This RG is applicable to licensees seeking changes in Not Applicable tic Risk Assessment in Risk- licensing basis, and is the responsibility of the Informed Decisions on Plant-Spe- licensee. cific Changes to the Licensing Basis 5 An Approach for Plant-Specific, - Not Applicable This RG is applicable to licensees seeking change to Not Applicable Risk Informed Decision making: licensing basis using a risk-informed approach and is Inservice Testing the responsibility of the COL applicant or licensee. NuScale is not using a risk-informed approach for IST. 7 An Approach for Plant-Specific, 1 Not Applicable This RG applies to existing licensees seeking NRC 16.1 Risk-Informed Decision making: approval of changes to their plant-specific technical Technical Specifications specifications. However, NuScale considered this guidance, as appropriate, in risk-informed technical specification development. 8 An Approach for Plant-Specific 1 Not Applicable This RG addresses the use of PRA in support of a risk- Not Applicable Risk-Informed Decision making informed inservice inspection program for piping. Conformance with Regulatory Criteria for Inservice Inspection of Piping Such a program is a plant-specific operational pro-gram that is the responsibility of the COL applicant or licensee. 9 Standard Format and Content of 1 Not Applicable This guidance governs site-specific decommission- Not Applicable License Termination Plans for ing and license termination planning and imple-Nuclear Power Reactors mentation activities that are the responsibility of the COL applicant or licensee.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 0 Guidelines for Evaluating Electro- 1 Partially Conforms Aspects of this guidance related to the design of SSC 3.11 magnetic and Radio-Frequency to address effects of electromagnetic and radio-fre- 7.2 Interference in Safety-Related quency interference (EMI/RFI) are applicable. 9.5 Instrumentation and Control Sys- Aspects of this guidance related to the design of tems site-specific SSC and installation and testing prac-tices for addressing the effects of EMI/RFI and power surges on safety-related I&C systems are the respon-sibility of the COL applicant or licensee. 1 Content of the Updated Final - Not Applicable This guidance governs site-specific reporting activi- Not Applicable Safety Analysis Report in Accor- ties that are the responsibility of the COL applicant dance with 10 CFR 50.71(e) or licensee. 3 Alternative Radiological Source 0 Partially Conforms For the NuScale design, the safety analysis shows 6.4 Terms for Evaluating Design Basis that core damage does not occur during any design 9.3 Accidents at Nuclear Power Reac- basis event. Thus, the RG 1.183 guidance is partially 12.2 tors applicable to the NuScale dose consequence analy-sis. The basis and justification for departures from 15.0.2 the RG 1.183 guidance for the limiting dose conse- 15.0.3 quence analysis for NuScale are provided in a Topi- 15.6 cal Report. NuScale uses the alternative source term non-LOCA or transient-specific guidance of RG 1.183 15.7 for Chapter 15 events. 15.10 4 Decommissioning of Nuclear 1 Not Applicable This RG governs site-specific decommissioning plan- Not Applicable Power Reactors ning and implementation activities that are the responsibility of the COL applicant or licensee. 5 Standard Format and Content for 1 Not Applicable This RG governs site-specific decommissioning plan- Not Applicable Post-Shutdown Decommission- ning activities that are the responsibility of the COL Conformance with Regulatory Criteria ing Activities Report applicant or licensee. 6 Guidance and Examples for Iden- - Not Applicable This RG endorses NEI 97-04 Appendix B is the Not Applicable tifying 10 CFR 50.2 Design Bases responsibility of the COL applicant or licensee. 7 Guidance for Implementation of - Not Applicable This RG implements change process requirements Not Applicable 10 CFR 50.59, Changes, Tests, and that are the responsibility of the COL applicant or Experiments licensee. 8 Standard Format and Content for 1 Not Applicable This RG is applicable to operating reactor licensees Not Applicable Applications to Renew Nuclear seeking to renew their operating licenses. Power Plant Operating Licenses

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 9 Fire Protection for Nuclear Power 2 Partially Conforms This RG is applicable except for portions (1) directed 9.4 Plants toward a specific reactor design (e.g., BWR or non- 9.5 LWR) or SSC conditions not relevant to the NuScale Appendix 9A PWR design; and (2) related to site-specific fire pro-tection systems and equipment or programmatic 11.3 and procedural activities that are the responsibility 14.3 of the COL applicant or licensee. 19.1 0 Calculational and Dosimetry - Partially Conforms The neutron flux and fluence calculation methods 5.3 Methods for Determining Pres- are consistent with the guidance of RG 1.190 with sure Vessel Neutron Fluence exceptions as described in NuScale Technical Report TR-0116-20781, "Fluence Calculation Methodology and Results." 1 Fire Protection Program for - Not Applicable This RG governs site-specific fire protection program Not Applicable Nuclear Power Plants During activities that are applicable only to holders of reac-Decommissioning and Perma- tor licenses that have permanently ceased power nent Shutdown operations. 2 Operation and Maintenance 1 Not Applicable Implementation of inservice testing is the responsi- Not Applicable Code Case Acceptability, ASME bility of the COL applicant or licensee. OM Code 3 ASME Code Cases Not Approved 4 Conforms ASME code cases in RG 1.193 are not used unless 5.2 for Use authorized by the NRC pursuant to 10 CFR 50.55a(z). 4 Atmospheric Relative Concentra- - Conforms None. 9.4 tions for Control Room Radiologi- 15.0.3 cal Habitability Assessments at Nuclear Power Plants 5 Methods and Assumptions for - Not Applicable This RG pertains to TID14844 source terms used by Not Applicable Conformance with Regulatory Criteria Evaluating Radiological Conse- licensees of existing reactors that are not authorized quences of Design Basis Acci- to use the alternative source term under dents at Light-Water Nuclear 10 CFR 50.67. Therefore, RG 1.183 is specified to be Power Reactors used in lieu of RG 1.195 for new reactors and existing reactors authorized to use the alternative source term under 10 CFR 50.67.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 6 Control Room Habitability at 1 Partially Conforms Aspects of this RG related to control room habitabil- 3.8 Light-Water Nuclear Power Reac- ity design within the scope of the standard plant 6.4 tors design are applicable to the DCA. References to ESF 18.7 (via the ventilation systems are not applicable to the NuS- Human-System cale design. The NuScale control room habitability Interface Design Results system neither relies on nor uses emergency filtra- Summary Report) tion to protect operators during accident conditions. Rather, clean air is provided using compressed air tanks. Other aspects of this guidance specify opera-tional, programmatic activities that are the responsi-bility of the COL applicant or licensee. 7 Demonstrating Control Room - Not Applicable Inleakage testing activities are the responsibility of Not Applicable Envelope Integrity at Nuclear the COL applicant or licensee referencing the certi-Power Reactors fied design. 8 Procedures and Criteria for - Not Applicable The evaluation governed by this guidance is the Not Applicable Assessing Seismic Soil Liquefac- responsibility of the COL applicant or licensee refer-tion at Nuclear Power Plant Sites encing the certified design. 9 Anchoring Components and - Partially Conforms The intent of this guidance is applicable but the spe- 3.8 Structural Supports in Concrete cific language endorses Appendix B of ACI 349-2001 with specified exceptions in the area of load combi-nations. NuScale uses the 2006 version of the ACI 349 standard. 0 An Approach for Determining the 2 Conforms As referenced in SRP 19.0 with regard to PRA quality 19.1 Technical Adequacy of Probabilis- and technical adequacy. tic Risk Assessment Results for Risk-Informed Activities Conformance with Regulatory Criteria

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 1 Guidelines for Categorizing Struc- 1 Partially Conforms 10 CFR 50.69 provides an alternative regulatory 3.2 tures, Systems, and Components framework for a licensee to use a risk-informed pro-in Nuclear Power Plants Accord- cess for categorizing SSC by their safety significance, ing to Their Safety Significance and based on this process can remove SSC of low safety significance from the scope of identified spe-cial treatment requirements. Thus, these require-ments are applicable to licensees that choose this alternative framework. NuScale uses a risk-informed, performance-based approach to safety classification that blends the strengths of deterministic engineer-ing judgment and probabilistic methods. Specifi-cally, the NuScale approach to SSC safety classification combines the traditional approach using the definitions of 10 CFR 50.2 and guidance of RG 1.26 and SRP Section 3.2.2 with the alternative regulatory framework similar to that prescribed in 10 CFR 50.69 and RG 1.201 (and NEI 00-04 endorsed by RG 1.201). This methodology is consistent with SECY-03-0047 and SECY-10-0034, which recom-mend the use of a probabilistic, risk-informed approach for SSC safety classification. NuScale applies the guidance of RG 1.201 and NEI 00-04 to the extent appropriate given the baseline risk met-rics for the NuScale advanced reactor design. 2 Standard Format and Content of - Not Applicable This RG implements regulatory requirements for Not Applicable Decommissioning Cost Estimates decommissioning cost estimates that are applicable for Nuclear Power Reactors only to licensees. Conformance with Regulatory Criteria 3 Transient and Accident Analysis - Conforms None. 15.0.2 Methods 4 Guidelines for Lightning Protec- - Conforms None. 3.8 tion of Nuclear Power Plants 7.0 7.2 8.1 8.3

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 5 Risk-Informed, Performance- 1 Not Applicable This RG applies to reactor licensees or applicants Not Applicable Based Fire Protection for Existing that are developing or revising a risk-informed, per-Light-Water Nuclear Power Plants formance-based fire protection program pursuant to 10 CFR 50.48(c). Development and implementa-tion of a risk-informed, performance-based fire pro-tection program would be the responsibility of COL applicants or licensees that reference the NuScale design, and that elect to implement the provisions of 10 CFR 50.48(c). 6 Combined License Applications - Partially Conforms This RG is the template for the FSAR layout, with Ch. 1 through Ch. 19 for Nuclear Power Plants (LWR exceptions. Edition) 7 Guidelines for Evaluating Fatigue - Conforms None. 3.8 Analyses Incorporating the Life 3.9 Reduction of Metal Components 3.12 Due to the Effects of the Light-Water Reactor Environment for New Reactors 8 A Performance-Based Approach - Not Applicable This guidance is for development of site-specific Not Applicable to Define the Site-Specific Earth- ground motion response spectra and is the respon-quake Ground Motion sibility of the COL applicant or licensee. 9 Guidelines for Environmental - Conforms None. 3.11 Qualification of Safety-Related Appendix 3C Computer-Based Instrumenta-7.2 tion and Control Systems in Nuclear Power Plants 14.3 Conformance with Regulatory Criteria 0 Qualification of Safety-Related - Not Applicable The NuScale design does not use safety-related bat- Not Applicable Battery Chargers and Inverters for tery chargers or inverters; EDSS battery chargers are Nuclear Power Plants not located in a harsh environment. 1 Qualification of Safety-Related - Conforms None. 3.11 Cables and Field Splices for Nuclear Power Plants 2 Sizing of Large Lead-Acid Stor- November 2008 Partially Conforms The NuScale DC power systems conform to the VRLA 8.3 age Batteries sizing guidance in IEEE Std. 485-1997, with consider-ation as appropriate for regulatory positions in RG 1.212 relevant to VRLA battery sizing.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 3 Qualification of Safety-Related - Not Applicable The NuScale electrical system design does not use Not Applicable Motor Control Centers for Nuclear safety-related motor control centers. Power Plants 4 Response Strategies for Potential 1 Conforms None. 19.5 Aircraft Threats 5 Guidance for ITAAC Closure 2 Not Applicable This guidance describes acceptable methods of Not Applicable Under 10 CFR Part 52 complying with the requirements of 10 CFR 52.99, which is applicable to COL applicants and licensees. 6 Containment Structural Integrity 0 Conforms None. 3.8 Evaluation for Internal Pressure 6.2 Loadings Above Design-Basis Pressure 7 Guidance for the Assessment of - Conforms None. 19.5 Beyond-Design-Basis Aircraft Impacts 8 Condition-Monitoring Tech- - Not Applicable The COL holder determines whether a cable is sub- Not Applicable niques for Electric Cables Used in ject to condition monitoring during the develop-Nuclear Power Plants ment of the maintenance rule (10 CFR 50.65) program. Cables that meet the criteria for inclusion in the maintenance rule program are subject to the guidance of RG 1.218. 9 Guidance on Making Changes to - Not Applicable These requirements are applicable to operating Not Applicable Emergency Plans for Nuclear reactor licensees, including COL holders. Power Reactors 1 Design-Basis Hurricane and Hurri- - Conforms NuScale uses the highest wind speed postulated in 3.3 cane Missiles for Nuclear Power Regulatory Position 1 (which occurs in Figure 2 of RG 3.5 Conformance with Regulatory Criteria Plants 1.221 Rev. 0) as the wind speed for the design basis 3.8 hurricane. Confirmation of site characteristics is the responsibility of the COL applicant. Radiological Environmental Mon- 2 Not Applicable This guidance governs site-specific, programmatic Not Applicable itoring for Nuclear Power Plants environmental monitoring activities that are the responsibility of the COL applicant or licensee. Preparation of Environmental 2 Not Applicable This guidance governs site-specific environmental Not Applicable Reports for Nuclear Power Sta- evaluation activities that are the responsibility of a tions license or construction permit applicant.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section 1 Supplement 1 to RG 4.2, Prepara- 1 Not Applicable This guidance is applicable to licensees seeking Not Applicable tion of Supplemental Environ- renewal of their operating license. mental Reports for Applications to Renew Nuclear Power Plant Operating Licenses Reporting Procedure for Mathe- - Not Applicable This guidance governs site-specific environmental Not Applicable matical Models Selected to Pre- activities related to modeling temperature impact of dict Heated Effluent Dispersion in plant discharge on aquatic systems. These activities Natural Water Bodies are the responsibility of the COL applicant or licensee. General Site Suitability Criteria for 2 Not Applicable This guidance governs site-specific evaluation activi- Not Applicable Nuclear Power Stations ties that are the responsibility of the COL applicant. Preparation of Environmental 1 Not Applicable This guidance applies only to uranium enrichment Not Applicable Reports for Commercial Uranium facilities. Enrichment Facilities Terrestrial Environmental Studies 2 Not Applicable This guidance governs site-specific environmental Not Applicable for Nuclear Power Stations evaluation activities that are the responsibility of a license or construction permit applicant. Performance, Testing, and Proce- 1 Not Applicable This guidance governs site-specific procedural activ- Not Applicable dural Specifications for Thermolu- ities that are the responsibility of a COL applicant or minescence Dosimetry: holder. Environmental Applications Radiological Effluent and Environ- 1 Not Applicable This guidance is applicable only to uranium mills. Not Applicable mental Monitoring at Uranium Mills Quality Assurance for Radiologi- 2 Not Applicable Applicable to COL applicants. Not Applicable Conformance with Regulatory Criteria cal Monitoring Programs (Incep-tion through Normal Operations to License Termination) - Effluent Streams and the Environment Monitoring and Reporting Radio- 2 Not Applicable This guidance is applicable only to fuel cycle facili- Not Applicable active Materials in Liquid and ties. Gaseous Effluents from Nuclear Fuel Cycle Facilities Standard Format and Content of 1 Not Applicable This guidance is applicable only to geological repos- Not Applicable Site Characterization Plans for itories. High-Level-Waste Geologic Repositories

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Standard Format and Content of - Not Applicable This guidance is applicable only to near-surface dis- Not Applicable Environmental Reports for Near- posal of radioactive waste, which is not within the Surface Disposal of Radioactive scope of the NuScale certified design. Waste Guidance for Selecting Sites for - Not Applicable This guidance is applicable only to near-surface dis- Not Applicable Near-Surface Disposal of Low- posal of radioactive waste, which is not within the Level Radioactive Waste scope of the NuScale certified design. Constraint on Releases of Air- 1 Not Applicable This guidance is applicable only to non-reactor facili- Not Applicable borne Radioactive Materials to ties. the Environment for Licensees other than Power Reactors Minimization of Contamination - Partially Conforms This guidance is applicable, except for the portions 9.1 and Radioactive Waste Genera- that relate to site-specific, operational aspects that 9.2 tion: Life-Cycle Planning are the responsibility of the COL applicant or 9.4 licensee referencing the NuScale design. 10.4 11.2 11.3 12.3 12.5 14.3 Decommissioning Planning - Not Applicable This RG is applicable to operating reactor licensees. Not Applicable During Operations Statistical Terminology and Nota- - Not Applicable This RG is directed towards licensees of fuel process- Not Applicable tion for Special Nuclear Materials ing and fuel fabrication facilities. Conformance with Regulatory Criteria Control and Accountability Standard Analytical Methods for - Not Applicable This RG is directed towards licensees of enrichment Not Applicable the Measurement of Uranium facilities. Tetrafluoride (UF4) and Uranium Hexafluoride (UF6) Standard Methods for Chemical, - Not Applicable This RG is directed towards licensees of fuel fabrica- Not Applicable Mass Spectrometric, and Spectro- tion facilities. chemical Analysis of Nuclear-Grade Uranium Dioxide Powders and Pellets

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Entry/Exit Control for Protected 1 Partially Conforms Site-specific, programmatic aspects of this guidance 13.6 (via Security Tech-Areas, Vital Areas, and Material are the responsibility of the COL applicant or nical Report) Access Areas licensee referencing the NuScale design. Design Considerations for Mini- 1 Not Applicable Applicable to Part 70 facilities. Not Applicable mizing Residual Holdup of Special Nuclear Material in Drying and Fluidized Bed Operations Guidelines for Germanium Spec- 2 Not Applicable This guidance governs processes, procedures, Not Applicable troscopy Systems for Measure- equipment, and methods that are not applicable to ment of Special Nuclear Material the NuScale design. Nondestructive Assay of Special 1 Not Applicable This RG applies to facilities that process SNM. The Not Applicable Nuclear Material Contained in NuScale design does not process SNM. Scrap and Waste General Use of Locks in the Pro- - Partially Conforms Site-specific, programmatic aspects of this guidance 13.6 (via Security Tech-tection and Controls of Facilities are the responsibility of the COL applicant or nical Report) and Special Nuclear Materials licensee referencing the NuScale design. Conduct of Nuclear Material Phys- - Not Applicable Applicable to Part 70 facilities. Not Applicable ical Inventories Limit of Error Concepts and Prin- - Not Applicable This RG is applicable to a special nuclear material Not Applicable ciples of Calculation in Nuclear licensee. Materials Control Training, Equipping, and Qualify- - Not Applicable Applicable to COL applicant or licensee. Not Applicable ing of Guards and Watchmen Nondestructive Uranium-235 1 Not Applicable Applicable to Part 70 facilities. Not Applicable Enrichment Assay by Gamma Ray Spectrometry Conformance with Regulatory Criteria Assessment of the Assumption of - Not Applicable This RG is not applicable to the DCA because NuS- Not Applicable Normality (Employing Individual cale is not a special nuclear material licensee. Observed Values) In Situ Assay of Plutonium Resid- 1 Not Applicable Applicable to Part 70 facilities. Not Applicable ual Holdup Design Considerations for Mini- - Not Applicable Applicable to Part 70 facilities. Not Applicable mizing Residual Holdup of Special Nuclear Material in Equipment for Wet Process Operations Selection of Material Balance 1 Not Applicable Applicable to Part 70 facilities. Not Applicable Areas and Item Control Areas

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Special Nuclear Material Door- - Not Applicable Applicable to COL applicant. Not Applicable way Monitors Evaluation of Shipper-Receiver - Not Applicable This RG applies to fuel processing and fuel fabrica- Not Applicable Differences in the Transfer of Spe- tion licensees. cial Nuclear Materials Nuclear Material Control systems 2 Not Applicable This RG is not applicable to the NuScale design but Not Applicable for Nuclear Power Plants may be used by a COL applicant to meet the mate-rial control and accounting requirements in Subpart B of 10 CFR Part 74. Specially Designed Vehicle with 1 Not Applicable Applicable to COL applicant. Not Applicable Armed Guards for Road Shipment of Special Nuclear Material Statistical Evaluation of Material - Not Applicable Applicable to Part 70 facilities. Not Applicable Unaccounted For Nondestructive Assay for Pluto- 1 Not Applicable Applicable to Part 70 processing. Not Applicable nium in Scrap Material by Sponta-neous Fission Detection Recommended Practice for Deal- 1 Not Applicable This RG is applicable to a special nuclear material Not Applicable ing with Outlying Observations licensee. General Methods for the Analysis - Not Applicable This RG is not applicable to the DCA because NuS- Not Applicable of Uranyl Nitrate Solutions for cale is not an applicant for a special nuclear material Assay, Isotopic Distribution, and in an unsealed form license. Impurity Determinations Design Considerations for Mini- - Not Applicable Applicable to Part 70 facilities. Not Applicable mizing Residual Holdup of Special Nuclear Material in Equipment for Conformance with Regulatory Criteria Dry Process Operations Plant Security Force Duties - Not Applicable Applicable to COL applicant or licensee. Not Applicable Perimeter Intrusion Alarm Sys- 3 Partially Conforms Site-specific, programmatic aspects of this guidance 13.6 (via Security Tech-tems are the responsibility of the COL applicant referenc- nical Report) ing the NuScale design. Design Considerations-Systems - Not Applicable Applicable to COL applicant or licensee. Not Applicable for Measuring the Mass of Liquids Internal Transfers of Special - Not Applicable Issued for comment. Not Applicable Nuclear Material (for Comment)

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Management Review of Nuclear 0 Not Applicable This RG applies to fuel cycle facilities. Not Applicable Material Control and Accounting Systems Standard Format and Content of 3 Not Applicable Not applicable to nuclear power plants. Not Applicable a Licensee Physical Protection Plan for Strategic Special Nuclear Material at Fixed Sites (Other than Nuclear Power Plants) Qualification, Calibration, and 1 Not Applicable This RG applies to fuel processing licensees. Not Applicable Error Estimation Methods for Nondestructive Assay Standard Format and Content of 1 Not Applicable This guidance governs site-specific physical protec- Not Applicable Safeguards Contingency Plans for tion features and security program activities that are Nuclear Power Plants the responsibility of the COL applicant or licensee. Standard Format and Content of 0 Not Applicable Applicable to fuel cycle facilities. Not Applicable Safeguards Contingency Plans for Fuel Cycle Facilities Standard Format and Content of 0 Not Applicable Applicable to transportation of special nuclear mate- Not Applicable Safeguards Contingency Plans for rial. Transportation Shipping and Receiving Control 1 Not Applicable Applicable to COL applicant or licensee. Not Applicable of Strategic Special Nuclear Mate-rial Considerations for Establishing 1 Not Applicable Applicable to COL applicant or licensee. Not Applicable Traceability of Special Nuclear Material Accounting Measure-Conformance with Regulatory Criteria ments Standard Format and Content for 1 Not Applicable Applicable to COL applicant or licensee. Not Applicable a Licensee Physical Security Plan for the Protection of Special Nuclear Material of Moderate or Low Strategic Significance Standard Format and Content of - Not Applicable Applicable to COL applicant or licensee. Not Applicable a Licensee Physical Protection Plan for Strategic Special Nuclear Material in Transit

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Intent and Scope of the Physical - Not Applicable This guidance applies to fuel cycle facilities. Not Applicable Protection Upgrade Rule Require-ments for Fixed Sites Reporting of Safeguards Events 1 Not Applicable This guidance applies to site-specific security issues Not Applicable concerning SNM and is the responsibility of the COL applicant or licensee. Physical Protection for Transient - Not Applicable Applicable to COL applicant or licensee. Not Applicable Shipments Vital Area Access Controls, Pro- - Partially Conforms Site-specific, programmatic aspects of this guidance 13.6 (via Security Tech-tection of Physical Security Equip- are the responsibility of the COL applicant or nical Report) ment, and Key and Lock Controls licensee referencing the NuScale design. Access Authorization Program for 2 Not Applicable This guidance governs site-specific physical security Not Applicable Nuclear Power Plants program activities that are the responsibility of the COL applicant or licensee. Protection Against Malevolent - Partially Conforms Although applicable to the COL applicant or 13.6 (via Security Tech-Use of Vehicles at Nuclear Power licensee, the design must allow compliance (e.g., nical Report) Plants F090 Site Layout Plan, which references parts of 73.55). Guidance for the Application of - Not Applicable This guidance is the responsibility of the COL appli- 13.6 (via Security Tech-Radiological Sabotage Design- cant or licensee. nical Report) Basis Threat in the Design, Devel-opment and Implementation of a Physical Security Program that Meets 10 CFR 73.55 Require-ments (SGI) Guidance for the Application of - Not Applicable COL applicant or licensee responsibility. Not Applicable Conformance with Regulatory Criteria the Theft and Diversion Design-Basis Treat in the Design Develop-ment, and Implementation of a Physical Security Program that Meets CFR 73.45 and 73.46 (SGI) Cyber Security Programs for - Partially Conforms The portions of RG 5.71 that govern site-specific 13.6 (via Security Tech-Nuclear Facilities operational and programmatic activities (e.g., devel- nical Report) opment and implementation of operational cyber security plans) apply to the COL applicant or licensee.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Fatigue Management for Nuclear - Not Applicable This RG is not applicable to the NuScale design but Not Applicable Power Plant Personnel may be used by a COL applicant or licensee to meet the fatigue management requirements of 10 CFR 26 Subpart I. Managing the Safety/Security - Not Applicable COL applicant or licensee responsibility. Not Applicable Interface Training and Qualification of - Not Applicable COL applicant or licensee responsibility. Not Applicable Security Personnel at Nuclear Power Reactor Facilities Physical Protection Programs at - Not Applicable This guidance governs site-specific physical protec- 13.6 (via Security Tech-Nuclear Power Reactors tion program activities that are the responsibility of nical Report) the COL applicant or licensee. Insider Mitigation Program (OUO- - Not Applicable COL applicant or licensee responsibility. Not Applicable SRI) Physical Protection of Mixed - Not Applicable The NuScale design does not use mixed oxide fuels. Not Applicable Oxide Fuels in Nuclear Power Plants (SGI) Protection of Safeguard Informa- - Conforms NuScale protects Safeguards Not Applicable tion Information against unauthorized disclosure in accordance with RG 5.79. Pressure-Sensitive and Tamper- - Not Applicable This RG is not applicable to the NuScale design. Not Applicable Indicating Device Seals for Mate-rial Control and Accounting of Special Nuclear Material Target Set Identification and - Not Applicable COL applicant or licensee responsibility. Not Applicable Development for Nuclear Power Conformance with Regulatory Criteria Reactors (OUO-SRI) Cyber Security Event Notifications - Not Applicable COL applicant or licensee responsibility. Not Applicable Fitness-For-Duty for New Nuclear - Not Applicable COL applicant or licensee responsibility. Not Applicable Power Plant Construction Sites Administrative Practices in Radia- 1 Not Applicable This guidance governs site-specific, programmatic Not Applicable tion Surveys and Monitoring activities related to radiation surveys and monitor-ing that are the responsibility of the COL applicant or licensee.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Personnel Monitoring Device - 1 Not Applicable This guidance governs site-specific programmatic Not Applicable Direct-Reading Pocket Dosime- activities related to the selection, maintenance, cali-ters bration, training, and reading of pocket dosimeters that are the responsibility of the COL applicant or licensee. Instructions for Recording and 2 Not Applicable This guidance governs site-specific programmatic Not Applicable Reporting Occupational Radia- activities related to recording and reporting dose tion Dose Data data that are the responsibility of the COL applicant or licensee. Information Relevant to Ensuring 3 Partially Conforms Implementation of this guidance is largely site-spe- 9.3 that Occupational Radiation cific and is the responsibility of the COL applicant. 10.4 Exposures at Nuclear Power Sta- However, the NuScale application for standard 11.2 tions Will Be as Low as Is Reason- design certification considered this guidance to be 11.4 ably Achievable applicable to the extent necessary to provide rea-sonable assurance that the COL applicant referenc- 11.5 ing the certified design can meet these 12.1 requirements. The aspects of this guidance that are 12.3 design-specific (i.e., pertain to design features, facili-ties, functions, and equipment that are technically 12.5 relevant to the NuScale standard plant design - e.g., 14.3 Position C.2) are applicable to the DCA. Acceptable Concepts, Models, 1 Not Applicable This guidance governs site-specific, programmatic Not Applicable Equations, and Assumptions for a activities, procedures, equipment, and methods that Bioassay Program are the responsibility of the COL applicant. Operating Philosophy for Main- 1-R Not Applicable These site-specific aspects are the responsibility of Not Applicable taining Occupational Radiation the COL applicant referencing the certified design. Conformance with Regulatory Criteria Exposures as Low as Is Reason-ably Achievable Applications of Bioassay for Ura- - Not Applicable This guidance governs programmatic activities that Not Applicable nium apply to licensees for which uranium bioassay is required. Instruction Concerning Prenatal 3 Not Applicable This guidance governs site-specific, programmatic Not Applicable Radiation Exposure activities, procedures, equipment, and methods that are the responsibility of the COL applicant. Acceptable Programs for Respira- 1 Not Applicable This guidance governs site-specific, programmatic Not Applicable tory Protection activities, procedures, equipment, and methods that are the responsibility of the COL applicant.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Information Relevant to Ensuring 2 Not Applicable This guidance governs activities applicable only to Not Applicable that Radiation Exposures at Medi- medical institutions. cal Institutions Will Be as Low as is Reasonably Achievable Occupational Radiation Dose 1 Partially Conforms This guidance is applicable, except for the portions 12.4 Assessment in Light Water Reac- that relate to site-specific, operational aspects that tor Power Plants - Design Stage are the responsibility of the COL applicant referenc-Man-Rem Estimates ing the NuScale design. Construction activities dose assessments are the responsibility of the COL appli-cant or licensee. Applications of Bioassay for 2 Not Applicable This guidance governs site-specific, programmatic Not Applicable Radioiodine activities, procedures, equipment, and methods that are the responsibility of the COL applicant or licensee. Health Physics Surveys for 1 Not Applicable Applicable only to processing and manufacturing Not Applicable Byproduct Material at NRC plants. Licensed Processing and Manu-facturing Plants Bioassay at Uranium Mills 1 Not Applicable Applicable only to uranium mills. Not Applicable Radiation Safety Surveys at Medi- 1 Not Applicable This guidance governs activities applicable only to Not Applicable cal Institutions medical institutions. Health Physics Surveys During 2 Not Applicable This guidance governs activities applicable only to Not Applicable Enriched Uranium-235 Processing facilities that process or fabricate fuel with uranium and Fuel Fabrication enriched with the U-235 isotope. Air Sampling in the Workplace 1 Not Applicable This guidance governs site-specific, programmatic Not Applicable activities related to air sampling in the workplace Conformance with Regulatory Criteria that are the responsibility of the COL applicant or licensee. Applications of Bioassay for Fis- - Not Applicable This guidance governs site-specific, programmatic Not Applicable sion and Activation Products activities, procedures, equipment, and methods that are the responsibility of the COL applicant or licensee. Radiation Protection Training for - Not Applicable This guidance governs site-specific operational Not Applicable Personnel at Light-Water-Cooled training programs, plans, and procedures that are Nuclear Power Plants the responsibility of the COL applicant or licensee.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Audible-Alarm Dosimeters - Not Applicable This guidance governs site-specific, programmatic Not Applicable activities, procedures, equipment, and methods that are the responsibility of the COL applicant or licensee. Instruction Concerning Risks from 1 Not Applicable This guidance governs site-specific, programmatic Not Applicable Occupational Radiation Exposure training and instructional activities that are the responsibility of the COL applicant or licensee. Health Physics Surveys in Ura- 1 Not Applicable This guidance governs activities applicable only to Not Applicable nium Recovery Facilities uranium recovery facilities. Information Relevant to Ensuring 1 Not Applicable This guidance governs activities applicable only to Not Applicable that Occupational Radiation uranium recovery facilities. Exposures at Uranium Recovery Facilities Will Be as Low as Is Rea-sonably Achievable Criteria for Establishing a Tritium - Not Applicable This guidance governs site-specific, programmatic Not Applicable Bioassay Program activities, procedures, equipment, and methods that are the responsibility of a licensee authorized to pos-sess nuclear material. Monitoring Criteria and Methods - Not Applicable This guidance governs site-specific, programmatic Not Applicable to Calculate Occupational Radia- activities, procedures, equipment, and methods that tion Doses are the responsibility of the COL applicant or licensee. Planned Special Exposure 1 Not Applicable This guidance governs site-specific, programmatic Not Applicable activities, procedures, equipment, and methods that are the responsibility of the COL applicant or licensee. Conformance with Regulatory Criteria Radiation Dose to the Embryo/ - Not Applicable This guidance governs site-specific, programmatic Not Applicable Fetus activities, procedures, equipment, and methods that are the responsibility of the COL applicant or licensee. ALARA Levels for Effluents from - Not Applicable This guidance governs activities applicable only to Not Applicable Materials Facilities materials facilities. Control of Access to High and 1 Partially Conforms Implementation of this guidance is site-specific and 12.1 Very High Radiation Areas in is the responsibility of the COL applicant. However, 12.3 Nuclear Power Plants NuScale considered this guidance to be applicable 12.5 to the extent necessary to provide reasonable assur-ance that the COL applicant or licensee referencing 14.2 the certified design can meet these requirements.

cale Final Safety Analysis Report RG Division Title Rev. Conformance Status Comments Section Release of Patients Administered - Not Applicable This guidance governs activities applicable only to Not Applicable Radioactive Materials facilities that administer radio-pharmaceuticals. Methods for Measuring Effective - Not Applicable This guidance governs dosimetry methods for deter- Not Applicable Dose Equivalent from External mining effective dose equivalent for external radia-Exposure tion exposures. These methods are the responsibility of the COL applicant or licensee. Conformance with Regulatory Criteria

cale Final Safety Analysis Report Standard (DSRS) P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 1.0, Rev 2: Introduction II.1 No Specific Acceptance Criteria - No Specific Acceptance Criteria. Not Applicable Interfaces 1.0, Rev 2: Introduction II.2 SRP Acceptance Criteria Associated Conforms None. Ch 1 Interfaces with Each Referenced SRP section 1.0, Rev 2: Introduction II.3 Performance of New Safety Features Conforms None. Ch 1 Interfaces and Design Qualification Testing Requirements 2.0, (March 2007): Site II.1 Specific SRP Acceptance Criteria Conforms This acceptance criterion is a pointer to 2.0 acteristics and Site Contained in Related SRP Chapter 2 other SRP sections. meters or Other Referenced SRP sections 2.0, (March 2007): Site II.2 COL Application Referencing an Early Not Applicable This acceptance criterion is applicable only Not Applicable acteristics and Site Site Permit to COL applicants that do not reference the meters DCA. 2.0, (March 2007): Site II.3 COL Application Referencing a Not Applicable This acceptance criterion is for COL Not Applicable acteristics and Site Certified Design applicants to meet the design parameters meters established in the DCA. 2.0, (March 2007): Site II.4 COL Application Referencing an Early Not Applicable This acceptance criterion is for COL Not Applicable acteristics and Site Site Permit and a Certified Design applicants to meet the design parameters meters established in the DCA. 2.0, (March 2007): Site II.5 COL Application Referencing Neither Not Applicable This acceptance criterion is applicable only Not Applicable acteristics and Site an Early Site Permit Nor a Certified to COL applicants that do not reference the meters Design DCA. 2.0, (March 2007): Site App A Table 1: Examples of Site Partially Conforms NuScale provides site parameters where Table 2.0-1 Conformance with Regulatory Criteria acteristics and Site Characteristics and Site Parameters applicable. meters 2.0, (March 2007): Site App A Table 2: Examples of Site-Related Partially Conforms NuScale provides site parameters where Table 2.0-1 acteristics and Site Design Parameters and Design applicable. meters Characteristics 2.1.1, Rev 3: Site Location All Specification of Location and Site Not Applicable Site-specific. Not Applicable Description Area Map 2.1.2, Rev 3: Exclusion All Establishment of Authority, Exclusion Not Applicable Site-specific. Not Applicable Authority and Control or Removal of Personnel and Property, and Proposed and Permitted Activities

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 2.1.3, Rev 3: Population All Population Data, Exclusion Area, Low- Not Applicable Site-specific. Not Applicable ibution Population Zone, Nearest Population Center Boundary, and Population Density 2.2.1-2.2.2, Rev 3: All Various Not Applicable Site-specific. Not Applicable tification of Potential rds in Site Vicinity 2.2.3, Rev 3: Evaluation of All Event Probability and Design-Basis Not Applicable Site-specific. Not Applicable ntial Accidents Event Analysis 2.3.1, Rev 3: Regional All Various Not Applicable Site-specific Not Applicable atology 2.3.1, Rev 3: Regional III.4.b.1 Postulated Site Parameters Conforms None. Table 2.0-1 atology 2.3.1 2.3.1, Rev 3: Regional III.4.b.2 Site Parameters Included as Tier 1 Conforms None. Table 2.0-1 atology Information 2.3.1 2.3.1, Rev 3: Regional III.4.b.3 Site Parameters Summary Table Conforms None. Table 2.0-1 atology 2.3.1 2.3.1, Rev 3: Regional III.4.b.4 Basis for Site Parameters Conforms None. Table 2.0-1 atology 2.3.1 2.3.2, Rev 3: Local II.1 thru II.4 Various Not Applicable Site-specific. Not Applicable orology 2.3.2, Rev 3: Local III.4.b.i Postulated Site Parameters Conforms None. Table 2.0-1 orology 2.3.2 2.3.2, Rev 3: Local III.4.b.ii Site Parameters Included as Tier 1 Conforms None. Table 2.0-1 Conformance with Regulatory Criteria orology Information 2.3.2 2.3.2, Rev 3: Local III.4.b.iii Site Parameters Summary Table Conforms None. Table 2.0-1 orology 2.3.2 2.3.2, Rev 3: Local III.4.b.iv Basis for Site Parameters Conforms None. Table 2.0-1 orology 2.3.2 2.3.3, Rev 3: Onsite All Various Not Applicable Site-specific. Not Applicable orological surements Program 2.3.4, Rev 3: Short-Term All Various Not Applicable Site-specific. Not Applicable ospheric Dispersion ates for Accident ases

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 2.3.4, Rev 3: Short-Term III.6.b.1 Postulated Site Parameters Conforms None. 2.3.4 ospheric Dispersion ates for Accident ases 2.3.4, Rev 3: Short-Term III.6.b.2 Site Parameters Included as Tier 1 Conforms None. 2.3.4 ospheric Dispersion Information ates for Accident ases 2.3.4, Rev 3: Short-Term III.6.b.3 Site Parameters Summary Table Conforms None. 2.3.4 ospheric Dispersion ates for Accident ases 2.3.4, Rev 3: Short-Term III.6.b.4 Basis for Site Parameters Conforms None. 2.3.4 ospheric Dispersion ates for Accident ases 2.3.4, Rev 3: Short-Term III (no number) Applicable Short-Term (Post- Conforms None. 2.3.4 ospheric Dispersion Accident) Site Parameters - EAB, LPZ, ates for Accident and Control Room Atmospheric ases Dispersion Factors 2.3.5, Rev 3: Long-Term All Various Not Applicable Site-specific. Not Applicable ospheric Dispersion ates for Routine ases 2.3.5, Rev 3: Long-Term III.5.b.1 Postulated Site Parameters Conforms None. 2.3.5 Conformance with Regulatory Criteria ospheric Dispersion ates for Routine ases 2.3.5, Rev 3: Long-Term III.5.b.2 Site Parameters Included as Tier 1 Conforms None. 2.3.5 ospheric Dispersion Information ates for Routine ases 2.3.5, Rev 3: Long-Term III.5.b.3 Site Parameters Summary Table Conforms None. 2.3.5 ospheric Dispersion ates for Routine ases

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 2.3.5, Rev 3: Long-Term III.5.b.4 Basis for Site Parameters Conforms None. 2.3.5 ospheric Dispersion ates for Routine ases 2.3.5, Rev 3: Long-Term III (no number) Applicable Long-Term (Routine Conforms None. 2.3.5 ospheric Dispersion Release) Site Parameters - Maximum ates for Routine Annual Average Site Boundary ases Atmospheric Dispersion Factor 2.4.1, Rev 3: Hydrologic All Various Not Applicable Site-specific. Not Applicable ription 2.4.1, Rev 3: Hydrologic III.7.B.i Postulated Site Parameters Conforms None. 2.4.1 ription 2.4.1, Rev 3: Hydrologic III.7.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.1 ription Information 2.4.1, Rev 3: Hydrologic III.7.B.iii Site Parameters Summary Table Conforms None. 2.4.1 ription 2.4.1, Rev 3: Hydrologic III.7.B.iv Basis for Site Parameters Conforms None. 2.4.1 ription 2.4.2, Rev 4: Floods All Various Not Applicable Site-specific. Not Applicable 2.4.2, Rev 4: Floods III.11.B.i Postulated Site Parameters Conforms None. 2.4.2 2.4.2, Rev 4: Floods III.11.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.2 Information 2.4.2, Rev 4: Floods III.11.B.iii Site Parameters Summary Table Conforms None. 2.4.2 2.4.2, Rev 4: Floods III.11.B.iv Basis for Site Parameters Conforms None. 2.4.2 Conformance with Regulatory Criteria 2.4.3, Rev 4: Probable All Various Not Applicable Site-specific. Not Applicable mum Flood (PMF) on ms and Rivers 2.4.3, Rev 4: Probable III.4.B.i Postulated Site Parameters Conforms None. 2.4.3 mum Flood (PMF) on ms and Rivers 2.4.3, Rev 4: Probable III.4.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.3 mum Flood (PMF) on Information ms and Rivers 2.4.3, Rev 4: Probable III.4.B.iii Site Parameters Summary Table Conforms None. 2.4.3 mum Flood (PMF) on ms and Rivers

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 2.4.3, Rev 4: Probable III.4.B.iv Basis for Site Parameters Conforms None. 2.4.3 mum Flood (PMF) on ms and Rivers 2.4.4, Rev 3: Potential All Various Not Applicable Site-specific. Not Applicable Failures 2.4.4, Rev 3: Potential III.8.B.i Postulated Site Parameters Conforms None. 2.4.4 Failures 2.4.4, Rev 3: Potential III.8.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.4 Failures Information 2.4.4, Rev 3: Potential III.8.B.iii Site Parameters Summary Table Conforms None. 2.4.4 Failures 2.4.4, Rev 3: Potential III.8.B.iv Basis for Site Parameters Conforms None. 2.4.4 Failures 2.4.5, Rev 3: Probable All Various Not Applicable Site-specific. Not Applicable mum Surge and Seiche ding 2.4.5, Rev 3: Probable III.7.B.i Postulated Site Parameters Conforms None. 2.4.5 mum Surge and Seiche ding 2.4.5, Rev 3: Probable III.7.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.5 mum Surge and Seiche Information ding 2.4.5, Rev 3: Probable III.7.B.iii Site Parameters Summary Table Conforms None. 2.4.5 mum Surge and Seiche Conformance with Regulatory Criteria ding 2.4.5, Rev 3: Probable III.7.B.iv Basis for Site Parameters Conforms None. 2.4.5 mum Surge and Seiche ding 2.4.6, Rev 3: Probable All Various Not Applicable Site-specific. Not Applicable mum Tsunami Hazards 2.4.6, Rev 3: Probable III.9.B.i Postulated Site Parameters Conforms None. 2.4.6 mum Tsunami Hazards 2.4.6, Rev 3: Probable III.9.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.6 mum Tsunami Hazards Information 2.4.6, Rev 3: Probable III.9.B.iii Site Parameters Summary Table Conforms None. 2.4.6 mum Tsunami Hazards

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 2.4.6, Rev 3: Probable III.9.B.iv Basis for Site Parameters Conforms None. 2.4.6 mum Tsunami Hazards 2.4.7, Rev 3: Ice Effects All Various Not Applicable Site-specific. Not Applicable 2.4.7, Rev 3: Ice Effects III.6.B.i Postulated Site Parameters Conforms None. 2.4.7 2.4.7, Rev 3: Ice Effects III.6.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.7 Information 2.4.7, Rev 3: Ice Effects III.6.B.iii Site Parameters Summary Table Conforms None. 2.4.7 2.4.7, Rev 3: Ice Effects III.6.B.iv Basis for Site Parameters Conforms None. 2.4.7 2.4.8, Rev 3: Cooling All Various Not Applicable Site-specific. Not Applicable er Canals and Reservoirs 2.4.8, Rev 3: Cooling III.5.B.1 Postulated Site Parameters Conforms None. 2.4.8 er Canals and Reservoirs 2.4.8, Rev 3: Cooling III.5.B.2 Site Parameters Included as Tier 1 Conforms None. 2.4.8 er Canals and Reservoirs Information 2.4.8, Rev 3: Cooling III.5.B.3 Site Parameters Summary Table Conforms None. 2.4.8 er Canals and Reservoirs 2.4.8, Rev 3: Cooling III.5.B.4 Basis for Site Parameters Conforms None. 2.4.8 er Canals and Reservoirs 2.4.9, Rev 3: Channel All Various Not Applicable Site-specific. Not Applicable rsions 2.4.9, Rev 3: Channel III.8.B.i Postulated Site Parameters Conforms None. 2.4.9 rsions 2.4.9, Rev 3: Channel III.8.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.9 rsions Information Conformance with Regulatory Criteria 2.4.9, Rev 3: Channel III.8.B.iii Site Parameters Summary Table Conforms None. 2.4.9 rsions 2.4.9, Rev 3: Channel III.8.B.iv Basis for Site Parameters Conforms None. 2.4.9 rsions 2.4.10, Rev 3: Flooding All Various Not Applicable Site-specific. Not Applicable ection Requirements 2.4.10, Rev 3: Flooding III.5.B.i Postulated Site Parameters Conforms None. 2.4.10 ection Requirements 2.4.10, Rev 3: Flooding III.5.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.10 ection Requirements Information

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 2.4.10, Rev 3: Flooding III.5.B.iii Site Parameters Summary Table Conforms None. 2.4.10 ection Requirements 2.4.10, Rev 3: Flooding III.5.B.iv Basis for Site Parameters Conforms None. 2.4.10 ection Requirements 2.4.11, Rev 3: Low Water All Various Not Applicable Site-specific. Not Applicable iderations 2.4.11, Rev 3: Low Water III.6.B.i Postulated Site Parameters Conforms None. 2.4.11 iderations 2.4.11, Rev 3: Low Water III.6.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.11 iderations Information 2.4.11, Rev 3: Low Water III.6.B.iii Site Parameters Summary Table Conforms None. 2.4.11 iderations 2.4.11, Rev 3: Low Water III.6.B.iv Basis for Site Parameters Conforms None. 2.4.11 iderations 2.4.12, Rev 3: II.1 thru II.5 Various Not Applicable Site-specific. Not Applicable ndwater 2.4.12, Rev 3: III.6.B.i Postulated Site Parameters Conforms None. 2.4.12 ndwater 2.4.12, Rev 3: III.6.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.12 ndwater Information 2.4.12, Rev 3: III.6.B.iii Site Parameters Summary Table Conforms None. 2.4.12 ndwater 2.4.12, Rev 3: III.6.B.iv Basis for Site Parameters Conforms None. 2.4.12 ndwater Conformance with Regulatory Criteria 2.4.13, Rev 3: Accidental All Various Not Applicable Site-specific. Not Applicable ases of Radioactive id Effluents in Ground Surface Waters 2.4.13, Rev 3: Accidental III.5.B.i Postulated Site Parameters Conforms None. 2.4.13 ases of Radioactive id Effluents in Ground Surface Waters 2.4.13, Rev 3: Accidental III.5.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.13 ases of Radioactive Information id Effluents in Ground Surface Waters

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 2.4.13, Rev 3: Accidental III.5.B.iii Site Parameters Summary Table Conforms None. 2.4.13 ases of Radioactive id Effluents in Ground Surface Waters 2.4.13, Rev 3: Accidental III.5.B.iv Basis for Site Parameters Conforms None. 2.4.13 ases of Radioactive id Effluents in Ground Surface Waters 2.4.14, Rev 3: Technical All Various Not Applicable Site-specific. Not Applicable ifications and rgency Operation irements 2.4.14, Rev 3: Technical III.5.B.i Postulated Site Parameters Conforms None. 2.4.14 ifications and rgency Operation irements 2.4.14, Rev 3: Technical III.5.B.ii Site Parameters Included as Tier 1 Conforms None. 2.4.14 ifications and Information rgency Operation irements 2.4.14, Rev 3: Technical III.5.B.iii Site Parameters Summary Table Conforms None. 2.4.14 ifications and rgency Operation irements 2.4.14, Rev 3: Technical III.5.B.iv Basis for Site Parameters Conforms None. 2.4.14 Conformance with Regulatory Criteria ifications and rgency Operation irements 2.5.1, Rev 4: Basic All Regional and Site Geology Not Applicable Site-specific. Not Applicable ogic and Seismic mation 2.5.2, Rev 4: Vibratory All Various Not Applicable Site-specific. Not Applicable nd Motion 2.5.2, Rev 4: Vibratory III.2.a Postulated Site Parameters Conforms None. 2.5.2 nd Motion

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 2.5.2, Rev 4: Vibratory III.2.b Site Parameters Included as Tier 1 Conforms None. 2.5 nd Motion Information 2.5.2, Rev 4: Vibratory III.2.c Site Parameters Summary Table Conforms None. 2.5.2 nd Motion 2.5.2, Rev 4: Vibratory III.2.d Basis for Site Parameters Conforms None. 2.5.2 nd Motion 2.5.3, Rev 4: Surface All Various Not Applicable Site-specific. Not Applicable ting 2.5.3, Rev 4: Surface III.2.a Postulated Site Parameters Conforms None. 2.5.3 ting 2.5.3, Rev 4: Surface III.2.b Site Parameters Included as Tier 1 Conforms None. 2.5.3 ting Information 2.5.3, Rev 4: Surface III.2.c Site Parameters Summary Table Conforms None. 2.5 ting 2.5.3, Rev 4: Surface III.2.d Basis for Site Parameters Conforms None. 2.5.3 ting 2.5.4, Rev 4: Stability of All Various Not Applicable Site-specific. Not Applicable urface Materials and dations 2.5.4, Rev 4: Stability of III.2.A Postulated Site Parameters Conforms None. 2.5.4 urface Materials and dations 2.5.4, Rev 4: Stability of III.2.B Site Parameters Included as Tier 1 Conforms None. 2.5.4 urface Materials and Information Conformance with Regulatory Criteria dations 2.5.4, Rev 4: Stability of III.2.C Site Parameters Summary Table Conforms None. 2.5 urface Materials and dations 2.5.4, Rev 4: Stability of III.2.D Basis for Site Parameters Conforms None. 2.5.4 urface Materials and dations 2.5.5, Rev 4: Stability of All Various Not Applicable Site-specific. Not Applicable es 2.5.5, Rev 4: Stability of III.2.A Postulated Site Parameters Conforms None. 2.5.5 es

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 2.5.5, Rev 4: Stability of III.2.B Site Parameters Included as Tier 1 Conforms None. 2.5.5 es Information 2.5.5, Rev 4: Stability of III.2.C Site Parameters Summary Table Conforms None. 2.5 es 2.5.5, Rev 4: Stability of III.2.D Basis for Site Parameters Conforms None. 2.5.5 es 3.2.1, Rev 2: Seismic II.1 Seismic Design Classification to Meet Partially Conforms This acceptance criterion is applicable 3.2.1 sification GDC 2; 10 CFR 100, Appendix A; and except that SSC meeting Staff Regulatory 10 CFR 50, Appendix S Guidance C.1.i of Regulatory Guide 1.29 are designated Seismic Category II rather than Seismic Category I. 3.2.2, Rev 2: System II.1 Quality Group Classification to Meet Conforms None. 3.2.2 ity Group Classification GDC 1 and 10 CFR 50.55a 3.2.2, Rev 2: System Table 3.2.21 Summary of Construction Codes and Partially Conforms This acceptance criterion is applicable Table 3.2-1 ity Group Classification Standards for Components of except for reference to RG 1.85, which was WaterCooled Nuclear Power Plants by withdrawn in 2004 because its guidance NRC Quality Classification System was updated and incorporated into RG 1.84. (Page 3.2.2-12) 3.2.2, Rev 2: System App. A and Additional Guidance for Classification Partially Conforms The intent of Table A-1 is applicable but Table 3.2-1 ity Group Classification Table A-1 of Systems and Components and some of the specific language refers to SSC Application of Quality Standards not part of the NuScale design. For example, the NuScale design does not include emergency diesel generators, ESF rooms, or pressurizer power operated relief valves. Conformance with Regulatory Criteria 3.3.1, Rev. 3: II.1 Most Severe Wind Partially Conforms Bounding parameters are established. 3.3.1 d Loadings 3.3.1, Rev. 3: II.2 Design Wind Speed, Recurrence Conforms None. 3.3.1 d Loadings Interval, and Other Site-Related Wind Parameters 3.3.1, Rev. 3: II.3 Procedures for Transforming Wind Conforms None. 3.3.1 d Loadings Speed Into Equivalent Pressure 3.3.2: Rev. 3: II.1 Most Severe Tornado Wind and Partially Conforms Bounding parameters are established. 3.3.2 ado Loads Associated Missiles 3.3.2: Rev. 3: II.2 Acceptance Criteria for Tornado Conforms None. 3.3.2 ado Loads Parameters and Spectrum of Tornado-Generated Missiles

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.3.2: Rev. 3: II.3 Procedures for Transforming Tornado Conforms None. 3.3.2 ado Loads Parameters Into Equivalent Loads on Structures 3.3.2: Rev. 3: II.4 Demonstrating That Failure of Conforms None. 3.3.2 ado Loads Structure or Component Not Designed for Tornado Loads Will Not Affect the Capability of Other SSC to Perform Safety Functions 3.4.1, Rev. 3: Internal II.1 Seismic Design and Classification Conforms None. 3.4.1 d Protection for Onsite Requirements pment Failures 3.4.1, Rev. 3: Internal II.2 Compliance with GDC 4 Conforms None. 3.4.1 d Protection for Onsite pment Failures 3.4.2, Rev. 3: Protection of II.1 Most Severe Highest Flood and Partially Conforms The NuScale Certified design assumes the 3.4.2 ctures Against Flood from Groundwater Levels NPP is above the maximum flood level. rnal Sources 3.4.2, Rev. 3: Protection of II.2 Highest Flood Level Below Grade - Conforms The NuScale Certified design assumes the 3.4.2 ctures Against Flood from Consideration of Hydrostatic Effects NPP is above the maximum flood level. rnal Sources and Wave Action 3.4.2, Rev. 3: Protection of II.3 Highest Flood Level Above Grade - Conforms The NuScale Certified design assumes the 3.4.2 ctures Against Flood from Consideration of Dynamic Loads NPP is above the maximum flood level. rnal Sources From Wave Action 3.5.1.1, Rev. 3: Internally- II.1 Statistical Significance of an Conforms None. 3.5.1 Conformance with Regulatory Criteria erated Missiles (Outside Identified Missile by Probability ainment) Analysis 3.5.1.1, Rev. 3: Internally- II.2 Acceptable Methods of Providing Conforms None. 3.5.1 erated Missiles (Outside Missile Protection ainment) 3.5.1.2, Rev. 3: Internally II.1 Statistical Significance of an Conforms None. 3.5.1 erated Missiles (Inside Identified Missile by Probability ainment) Analysis 3.5.1.2, Rev. 3: Internally II.2 Acceptable Methods of Providing Conforms None. 3.5.1 erated Missiles (Inside Missile Protection ainment)

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 3.5.1.3, Rev. 0: Turbine II.1 Probability of Unacceptable Damage Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable iles From Turbine Missiles Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 do not have to rely on the turbine missile generation probabilities, including turbine rotor integrity. S 3.5.1.3, Rev. 0: Turbine II.2 Turbine Missile Generation Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable iles Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 do not have to rely on the turbine missile generation probabilities, including turbine rotor integrity. S 3.5.1.3, Rev. 0: Turbine II.3 Acceptably Low Missile Generation Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable iles Probability Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 do not have to rely on the turbine missile generation probabilities, including turbine rotor integrity. S 3.5.1.3, Rev. 0: Turbine II.4 Missile Generation Probability Tables Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable iles From Turbine Manufacturers Section I.1, plants that use barriers to (Including Table 3.5.1.3-1) protect essential SSCs specified in RG 1.115 do not have to rely on the turbine missile generation probabilities, including turbine rotor integrity. S 3.5.1.3, Rev. 0: Turbine II.5 Inservice Inspection and Test Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable Conformance with Regulatory Criteria iles Program for Applicants Obtaining Section I.1, plants that use barriers to Turbine From Manufacturers without protect essential SSCs specified in RG 1.115 NRC-Approved Procedures for do not have to rely on the turbine missile Calculating Missile Generation generation probabilities, including turbine Probabilities rotor integrity. S 3.5.1.3, Rev. 0: Turbine II.6 Protective Barriers Conforms None. 3.5.1.3 iles 3.5.1.4, Rev. 4: Missiles II.1 Design Basis Tornado-Generated Conforms The NuScale design also includes RG 1.221 3.5.1.4 erated by Extreme Winds Missile Spectrum for Design Basis Hurricane-Generated Missiles. 3.5.1.4, Rev. 4: Missiles II.2 Statistical Significance of an Conforms None. 3.5.1.4 erated by Extreme Winds Identified Missile by Probability

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.5.1.4, Rev. 4: Missiles II.3 Identifying Appropriate Design Basis Conforms None. 3.5.1.4 erated by Extreme Winds Missiles Generated by Natural Phenomena 3.5.1.5, Rev 4: Site II.1 Compliance with 10 CFR 100 Not Applicable The NuScale design assumes no proximity Not Applicable imity Missiles (Except missiles. aft) 3.5.1.5, Rev 4: Site II.2 Compliance with GDC 4 Not Applicable The NuScale design assumes no proximity Not Applicable imity Missiles (Except missiles. aft) 3.5.1.6, Rev 4: Aircraft II.1 and II.2 Various Not Applicable The NuScale design assumes no aircraft Not Applicable rds hazard missiles. 3.5.1.6, Rev 4: Aircraft III.8.B.1 Postulated Site Parameters Conforms The NuScale design assumes no aircraft Table 2.0-1 rds hazard missiles. 3.5.1.6, Rev 4: Aircraft III.8.B.2 Site Parameters Included as Tier 1 Conforms The NuScale design assumes no aircraft Table 2.0-1 rds Information hazard missiles. 3.5.1.6, Rev 4: Aircraft III.8.B.3 Site Parameters Summary Table Conforms The NuScale design assumes no aircraft Table 2.0-1 rds hazard missiles. 3.5.1.6, Rev 4: Aircraft III.8.B.4 Basis for Site Parameters Conforms The NuScale design assumes no aircraft Table 2.0-1 rds hazard missiles. 3.5.2, Rev 3: Structures, II (no number) Capability of SSC to Withstand the Conforms None. 3.5.2 ems, and Components to Effects of Externally Generated rotected From Externally- Missiles erated Missiles 3.5.3, Rev. 3: Barrier II.1.A For Local Damage Prediction - Departure NuScale uses a finite element analysis for 3.5.3 Conformance with Regulatory Criteria gn Procedures Concrete predicting penetration distance of turbine missiles in concrete, rather than the modified National Defense Research Council (NDRC) formula specified in Section II.1.A. In some locations perforation and scabbing is predicted. However, given the physical separation of the redundant safety-related equipment, there is no turbine missile that can prevent essential systems from performing their function. 3.5.3, Rev. 3: Barrier II.1.B For Local Damage Prediction - Steel Conforms None. 3.5.3 gn Procedures

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.5.3, Rev. 3: Barrier II.1.C For Local Damage Prediction - Not Applicable This acceptance criterion specifies Not Applicable gn Procedures Composite sections provisions when using composite or multi-element barriers. NuScale does not use composite or multi-element barriers. 3.5.3, Rev. 3: Barrier II.2 For Overall Damage Prediction Partially Conforms This acceptance criterion is applicable 3.5.3 gn Procedures except for reference to subtier ANSI/AISC N690-1994 with Supplement 2 (2004). NuScale uses the 2012 version of this standard. 3.6.1, Rev 3: Plant Design II.1 Separation of High and Moderate Conforms None. 3.6.1 rotection Against Energy Fluid Systems From Essential ulated Piping Failures in Systems/Components Systems Outside ainment 3.6.1, Rev 3: Plant Design II.2 High and Moderate Energy Fluid Conforms None. 3.6.1 rotection Against Systems Are Enclosed ulated Piping Failures in Systems Outside ainment 3.6.1, Rev 3: Plant Design II.3 Cases Where Neither Physical Conforms None. 3.6.1 rotection Against Separation Nor Protective Enclosures ulated Piping Failures in Are Practical Systems Outside ainment 3.6.1, Rev 3: Plant Design II.4 Design Features Conforms None. 3.6.1 Conformance with Regulatory Criteria rotection Against ulated Piping Failures in Systems Outside ainment 3.6.1, Rev 3: Plant Design II.5 Effects of Postulated Failures Conforms None. 3.6.1 rotection Against ulated Piping Failures in Systems Outside ainment

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.6.2, Rev 2: II.1 Postulated Pipe Rupture Locations Conforms None. 3.6.2 rmination of Rupture Inside Containment tions and Dynamic ts Associated with the ulated Rupture of Piping 3.6.2, Rev 2: II.2 Postulated Pipe Rupture Locations Conforms None. 3.6.2 rmination of Rupture Outside Containment tions and Dynamic ts Associated with the ulated Rupture of Piping 3.6.2, Rev 2: II.3 Methods of Analysis Conforms None. 3.6.2 rmination of Rupture tions and Dynamic ts Associated with the ulated Rupture of Piping 3.6.2, Rev 2: III.1 Pipe Break Criteria Conforms None. 3.6.2 rmination of Rupture tions and Dynamic ts Associated with the ulated Rupture of Piping 3.6.2, Rev 2: III.2 Dynamic Effects Conforms None. 3.6.2 rmination of Rupture tions and Dynamic ts Associated with the ulated Rupture of Piping Conformance with Regulatory Criteria 3.6.2, Rev 2: III.3 Assumptions for Modeling Jet Conforms None. 3.6.2 rmination of Rupture Impingement Forces tions and Dynamic ts Associated with the ulated Rupture of Piping 3.6.2, Rev 2: III.4 Analyses of Pipe Break Dynamic Conforms None. 3.6.2 rmination of Rupture Effects on Mechanical Components tions and Dynamic and Supports ts Associated with the ulated Rupture of Piping 3.6.3, Rev 1: Leak-Before- II.1 Compliance with GDC 4 Conforms None. 3.6.3 k Evaluation Procedures

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.6.3, Rev 1: Leak-Before- II.2 Low Probability of Pipe Rupture Conforms None. 3.6.3 k Evaluation Procedures S 3.7.1, Rev. 0: Seismic II.1 Design Ground Motion Conforms None. 3.7.1 gn Parameters S 3.7.1, Rev. 0: Seismic II.2 Percentage of Critical Damping Conforms None. 3.7.1 gn Parameters Values S 3.7.1, Rev. 0: Seismic II.3 Supporting Media for Seismic Conforms None. 3.7.1 gn Parameters Category I Structures S 3.7.1, Rev. 0: Seismic II.4 Review Considerations for DC and Conforms None. 3.7.1 gn Parameters COL Applications S 3.7.2, Rev 0: Seismic II.1 Seismic Analysis Methods Conforms None. 3.7.2 em Analysis S 3.7.2, Rev 0: Seismic II.2 Natural Frequencies and Responses Conforms None. 3.7.2 em Analysis S 3.7.2, Rev 0: Seismic II.3 Procedures Used for Analytical Conforms None. 3.7.2 em Analysis Modeling S 3.7.2, Rev 0: Seismic II.4 Soil-Structure Interaction Conforms None. 3.7.2 em Analysis S 3.7.2, Rev 0: Seismic II.5 Development of In-Structure Conforms None. 3.7.2 em Analysis Response Spectra S 3.7.2, Rev 0: Seismic II.6 Three Components of Design Ground Conforms None. 3.7.2 em Analysis Motion S 3.7.2, Rev 0: Seismic II.7 Combination of Modal Responses Conforms None. 3.7.2 em Analysis Conformance with Regulatory Criteria S 3.7.2, Rev 0: Seismic II.8 Interaction of Non-Seismic Category I Conforms None. 3.7.2 em Analysis Structures with Seismic Category I SSCs S 3.7.2, Rev 0: Seismic II.9 Effects of Parameter Variations on Conforms None. 3.7.2 em Analysis Floor Response Spectra S 3.7.2, Rev 0: Seismic II.10 Use of Equivalent Vertical Static Conforms None. 3.7.2 em Analysis Factors S 3.7.2, Rev 0: Seismic II.11 Methods Used to Account for Conforms None. 3.7.2 em Analysis Torsional Effects S 3.7.2, Rev 0: Seismic II.12 Comparison of Responses Not Applicable NuScale does not perform both time history Not Applicable em Analysis analysis and response spectrum analysis in its analysis of structures.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 3.7.2, Rev 0: Seismic II.13 Analysis Procedure for Damping Conforms None. 3.7.1 em Analysis 3.7.2 S 3.7.2, Rev 0: Seismic II.14 Determination of Overturning Conforms None. 3.7.2 em Analysis Moments and Sliding Forces, Structure to Soil Pressures and Frictional Forces for Seismic Category I Structures S 3.7.3, Rev. 0: Seismic II.1 Seismic Analysis Methods Conforms None. 3.7.3 ystem Analysis S 3.7.3, Rev. 0: Seismic II.2 Determination of Number of Conforms None. 3.7.3 ystem Analysis Earthquake Cycles S 3.7.3, Rev. 0: Seismic II.3 Procedures Used for Analytical Conforms None. 3.7.3 ystem Analysis Modeling S 3.7.3, Rev. 0: Seismic II.4 Basis for Selection of Frequencies Conforms None. 3.7.3 ystem Analysis S 3.7.3, Rev. 0: Seismic II.5 Analysis Procedure for Damping Conforms None. 3.7.3 ystem Analysis S 3.7.3, Rev. 0: Seismic II.6 Three Components of Design Ground Conforms None. 3.7.3 ystem Analysis Motion S 3.7.3, Rev. 0: Seismic II.7 Combination of Modal Responses Conforms None. 3.7.3 ystem Analysis S 3.7.3, Rev. 0: Seismic II.8 Interaction of Non-Seismic Category I Conforms None. 3.7.3 ystem Analysis Subsystems with Seismic Category I SSCs Conformance with Regulatory Criteria S 3.7.3, Rev. 0: Seismic II.9 Multiply-Supported Equipment and Conforms None. 3.7.3 ystem Analysis Components with Distinct Inputs S 3.7.3, Rev. 0: Seismic II.10 Use of Equivalent Vertical Static Conforms None. 3.7.3 ystem Analysis Factors S 3.7.3, Rev. 0: Seismic II.11 Torsional Effects of Eccentric Masses Conforms None. 3.7.3 ystem Analysis S 3.7.3, Rev. 0: Seismic II.12 Seismic Category I Buried Piping, Conforms None. 3.7.3 ystem Analysis Conduits, and Tunnels S 3.7.3, Rev. 0: Seismic II.13 Methods for Seismic Analysis of Not Applicable The NuScale design does not use dams. Not Applicable ystem Analysis Seismic Category I Concrete Dams S 3.7.3, Rev. 0: Seismic II.14 Methods for Seismic Analysis of Conforms None. 3.7.3 ystem Analysis Above-Ground Tanks

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.7.4, Rev 2: Seismic II.1 Comparison with RG 1.12 Partially Conforms There is a COL item to comply. Locations are 3.7.4 umentation identified in conformance with RG 1.12, however seismic instrumentation cannot be placed inside containment. 3.7.4, Rev 2: Seismic II.2 Comparison with RG 1.166 Not Applicable See RG 1.166 in Table 1.9-2. Not Applicable umentation 3.7.4, Rev 2: Seismic II.3 Comparison with the requirements of Not Applicable Identified as an expectation for COL Not Applicable umentation 10 CFR 20.1101 (ALARA) applicants. 3.8.1, Rev 4: Concrete All Various Not Applicable The NuScale design does not have a Not Applicable ainment concrete containment. S 3.8.2, Rev. 0: Steel II.1 Description of the Containment Conforms None. 3.8.2 ainment S 3.8.2, Rev. 0: Steel II.2 Applicable Codes, Standards, and Conforms None. 3.8.2 ainment Specifications S 3.8.2, Rev. 0: Steel II.3 Loads and Loading Combinations Conforms None. 3.8.2 ainment S 3.8.2, Rev. 0: Steel II.4 Design and Analysis Procedures Conforms None. 3.8.2 ainment S 3.8.2, Rev. 0: Steel II.5 Structural Acceptance Criteria Conforms None. 3.8.2 ainment S 3.8.2, Rev. 0: Steel II.6 Materials, Quality Control, and Special Conforms None. 3.8.2 ainment Construction Techniques S 3.8.2, Rev. 0: Steel II.7 Testing and Inservice Surveillance Conforms None. 3.8.2 ainment Requirements Conformance with Regulatory Criteria 3.8.3, Rev 4: Concrete and All Various Not Applicable The NuScale containment does not have Not Applicable l Internal Structures of internal structures. l or Concrete ainments S 3.8.4, Rev. 0, Other II.1 Description of the Structures Conforms None. 3.8.4 mic Category I Structures S 3.8.4, Rev. 0, Other II.2 Applicable Codes, Standards, and Conforms None. 3.8.4 mic Category I Structures Specifications S 3.8.4, Rev. 0, Other II.3 Loads and Load Combinations Conforms None. 3.8.4 mic Category I Structures S 3.8.4, Rev. 0, Other II.3.A Concrete Structures Conforms None. 3.8.4 mic Category I Structures

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 3.8.4, Rev. 0, Other II.3.B Steel Structures Conforms None. 3.8.4 mic Category I Structures S 3.8.4, Rev. 0, Other II.4 Design and Analysis Procedures Conforms None. 3.8.4 mic Category I Structures S 3.8.4, Rev. 0, Other II.5 Structural Acceptance Criteria Conforms None. 3.8.4 mic Category I Structures S 3.8.4, Rev. 0, Other II.6 Materials, Quality Control, and Special Conforms None. 3.8.4 mic Category I Structures Construction Techniques S 3.8.4, Rev. 0, Other II.7 Testing and Inservice Surveillance Conforms None. 3.8.4 mic Category I Structures Requirements S 3.8.4, Rev. 0, Other II.8 Masonry Walls Not Applicable Masonry walls are not used in the NuScale Not Applicable mic Category I Structures design. S 3.8.5, Rev. 0: II.1 Description of the Foundation Conforms None. 3.8.5 dations S 3.8.5, Rev. 0: II.2 Applicable Codes, Standards, and Conforms None. 3.8.5 dations Specifications S 3.8.5, Rev. 0: II.3 Loads and Load Combinations Conforms None. 3.8.5 dations S 3.8.5, Rev. 0: II.4 Design and Analysis Procedures Conforms None. 3.8.5 dations S 3.8.5, Rev. 0: II.5 Structural Acceptance Criteria Conforms None. 3.8.5 dations S 3.8.5, Rev. 0: II.6 Materials, Quality Control, and Special Conforms None. 3.8.5 dations Construction Techniques Conformance with Regulatory Criteria S 3.8.5, Rev. 0: II.7 Testing and Inservice Surveillance Conforms None. 3.8.5 dations Requirements 3.9.1, Rev 3: Special II.1 Specification of Transients Conforms None. 3.9.1 cs for Mechanical ponents 3.9.1, Rev 3: Special II.2 Computer Programs to be Used in Conforms None. 3.9.1 cs for Mechanical Dynamic and Static Analyses ponents 3.9.1, Rev 3: Special II.3 Use of Experimental Stress Analysis Not Applicable Experimental Stress Analysis Method is not Not Applicable cs for Mechanical Methods in Lieu of Analytical used. ponents Methods

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.9.1, Rev 3: Special II.4 When Service Level D Limits are Conforms None. 3.9.1 cs for Mechanical Specified for Code Class 1 and Core ponents Support Components 3.9.2, Rev 3: Dynamic II.1 Vibration, Thermal Expansion, and Partially Conforms This acceptance criterion is applicable 3.9.2 ng and Analysis of Dynamic Effects Testing except for aspects related to test ems, Structures, and performance and associated corrective ponents actions (as required), which are the responsibility of the COL applicant referencing the certified design. 3.9.2, Rev 3: Dynamic II.2 Compliance with GDC 2 Conforms None. 3.9.2 ng and Analysis of ems, Structures, and ponents 3.9.2, Rev 3: Dynamic II.3 Analytical Solutions to Predict Conforms None. 3.9.2 ng and Analysis of Vibrations of Reactor Internals for ems, Structures, and Prototype Plants ponents 3.9.2, Rev 3: Dynamic II.4 Preoperational Vibration and Stress Conforms None. 3.9.2 ng and Analysis of Test Program ems, Structures, and ponents 3.9.2, Rev 3: Dynamic II.5 Structural Design Adequacy of Conforms None. 3.9.2 ng and Analysis of Reactor Internals and Reactor Coolant ems, Structures, and Piping ponents Conformance with Regulatory Criteria 3.9.2, Rev 3: Dynamic II.6 Correlation of Tests and Analyses of Conforms None. 3.9.2 ng and Analysis of Reactor Internals ems, Structures, and ponents 3.9.2, Rev 3: Dynamic II.7 Test Specifications for New Conforms None. 3.9.2 ng and Analysis of Applications ems, Structures, and ponents

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.9.3, Rev 3: ASME Code II.1 Loading Combinations, System Conforms None. 3.9.3 1, 2, and 3 Components Operating Transients, and Stress Component Supports, Limits Core Support Structures 3.9.3, Rev 3: ASME Code II.2 Design and Installation of Pressure Conforms None. 3.9.3 1, 2, and 3 Components Relief Devices Component Supports, Core Support Structures 3.9.3, Rev 3: ASME Code II.3 Component Supports Not Applicable NRC Bulletin 88-11 applies to PWR designs Not Applicable 1, 2, and 3 Components that incorporate a pressurizer separate from Component Supports, the reactor pressure vessel, with a surge line Core Support Structures connecting the two. In the NuScale design, the pressurizer is integral (i.e., is located within) to the reactor pressure vessel: there is no pressurizer surge line within which thermal stratification (that is the issue of this bulletin) would occur. 3.9.4, Rev 3: Control Rod II.1 Adequacy of Descriptive Information Conforms This acceptance criterion is applicable 3.9.4 e Systems (seismic design per RG 1.29) but contains a typographical error. The wording is confusing because it mixes an SRP section reference with a RG. 3.9.4, Rev 3: Control Rod II.2 Codes and Standards for Conforms None. 3.9.4 e Systems Construction 3.9.4, Rev 3: Control Rod II.3 Load Combination Sets for Design Conforms None. 3.9.4 Conformance with Regulatory Criteria e Systems and Service Conditions Defined in ASME Code Section III, NB-3113 3.9.4, Rev 3: Control Rod II.4 Operability Assurance Program Conforms None. 3.9.4 e Systems 3.9.5, Rev 3: Reactor II.1 Loads, Loading Combinations, and Conforms None. 3.9.3 sure Vessel Internals Limits for Portions Constructed to 3.9.5 ASME Code Section NG 3.9.5, Rev 3: Reactor II.2 Design and Construction of Core Conforms None. 3.9.3 sure Vessel Internals Support Structures 3.9.5

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.9.5, Rev 3: Reactor II.3 Design Criteria, Loading Conditions, Conforms None. 3.9.2 sure Vessel Internals and Analyses for Design of Reactor Internals Other Than Core Support Structures 3.9.5, Rev 3: Reactor II.4 Deformation Limits for Reactor Conforms None. 3.9.5 sure Vessel Internals Internals 3.9.5, Rev 3: Reactor II.5 Design of Reactor Internals to Partially Conforms The intent of subtier NUREG-0609 is 3.9.5 sure Vessel Internals Accommodate Asymmetric applicable but the language refers to a Blowdown Loads From Postulated different type of LWR and SSC conditions Pipe Ruptures not relevant to the NuScale design. Specifically, this guidance provides methodology for evaluation of loading transients and structural components, including containment subcompartment analysis, when a double-ended guillotine break of reactor coolant loop piping occurs at the reactor vessel inlet. The NuScale containment vessel design does not have subcompartments. In addition, the NuScale design does not have reactor coolant loops. Notwithstanding the above, this guidance is applicable to the evaluation of loading transients and structural components for postulated breaks of chemical and volume control system (CVCS) piping and piping at Conformance with Regulatory Criteria the reactor vent valves. 3.9.5, Rev 3: Reactor II.6 Effects of Flow-Induced Vibration and Partially Conforms This acceptance criterion (including 3.9.5 sure Vessel Internals Acoustic Resonances (Including Appendix A) is applicable except for aspects Appendix A) that are BWR-specific. 3.9.6, Rev 3: Functional II.1 Functional Design and Qualification Partially Conforms This acceptance criterion is applicable 3.9.6 gn, Qualification, and of Pumps, Valves, and Dynamic except for aspects related to functional vice Testing Programs Restraints design, qualification, and testing of safety-umps, Valves, and related pumps. Safety-related pumps are amic Restraints not used in the NuScale design.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.9.6, Rev 3: Functional II.2 Inservice Testing Program for Pumps Partially Conforms This acceptance criterion is applicable 3.9.6 gn, Qualification, and except for aspects related to inservice vice Testing Programs testing of safety-related pumps. Safety-umps, Valves, and related pumps are not used in the NuScale amic Restraints design. The only pumps that fall within the scope of this criterion in the NuScale design are the CVCS pumps. These pumps are ASME Class III because they contain reactor coolant during normal operation, but they serve no safety function. 3.9.6, Rev 3: Functional II.3 Inservice Testing Program for Valves Partially Conforms Refer to Section 3.9.6.3.2 for valve testing 3.9.6 gn, Qualification, and and Section 3.9.6.6 for augmented valves vice Testing Programs testing program. umps, Valves, and amic Restraints 3.9.6, Rev 3: Functional II.4 Inservice Testing Program for Not Applicable The NuScale Power Plant does not have Not Applicable gn, Qualification, and Dynamic Restraints pumps or dynamic restraints that perform a vice Testing Programs specific function identified in the ASME OM umps, Valves, and Code Subsection ISTA-1100. amic Restraints 3.9.6, Rev 3: Functional II.5 Relief Requests and Proposed Conforms Refer to Section 3.9.6.5 for relief requests 3.9.6 gn, Qualification, and Alternatives and alternative authorizations to the code. vice Testing Programs umps, Valves, and amic Restraints Conformance with Regulatory Criteria 3.9.6, Rev 3: Functional II.6 Operational Programs Not Applicable This acceptance criterion is related to Not Applicable gn, Qualification, and operational activities, including vice Testing Programs implementation of pre-service testing, umps, Valves, and inservice testing and inspection, and motor-amic Restraints operated valve testing programs, that are the responsibility of the COL applicant referencing the certified design.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3.9.7, Rev 0: Risk- All Various Not Applicable Development and implementation of a risk- Not Applicable med Inservice Testing informed, performance-based inservice testing program is the responsibility of COL applicants that reference the NuScale certified design and that elect to implement such a program. 3.9.8, Rev 0: Standard All Various Not Applicable Development and implementation of a risk- Not Applicable ew Plan for the Review of informed, inservice inspection program for Informed Inservice piping is the responsibility of COL ection of Piping applicants that reference the NuScale certified design, and that elect to implement such a program. 3.10, Rev 3: Seismic and II.1 Qualification of Electrical Equipment Conforms None. 3.10 amic Qualification of and Associated Supports hanical and Electrical pment 3.10, Rev 3: Seismic and II.2 Testing of Instrumentation Described Partially Conforms See RG 1.97 in Table 1.9-2 3.11 amic Qualification of in RG 1.97 hanical and Electrical pment 3.10, Rev 3: Seismic and II.3 Experience-Based Qualification Not Applicable Experience based seismic qualification is Not Applicable amic Qualification of not used. hanical and Electrical pment 3.10, Rev 3: Seismic and II.4 Records Conforms The NuScale design indicates that a Records 3.10 Conformance with Regulatory Criteria amic Qualification of program is required and includes a COL hanical and Electrical item to maintain one. pment 3.10, Rev 3: Seismic and II.5 Qualification Program for Valves that Conforms None. 3.10 amic Qualification of are Part of the Reactor Coolant hanical and Electrical Pressure Boundary pment 3.10, Rev 3: Seismic and II.6 Documentation of Qualification Conforms None. 3.10 amic Qualification of Program hanical and Electrical pment

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 3.11, Rev. 0: II.1 Application of RG 1.89 for Partially Conforms See RG 1.89 in Table 1.9-2. 3.11 ronmental Qualification Environmental Qualification Program echanical and Electrical per 10 CFR 50.49 pment S 3.11, Rev. 0: II.2 Application of Clarification Related to Conforms None. 3.11 ronmental Qualification IEEE Std. 323 Criteria echanical and Electrical pment S 3.11, Rev. 0: II.3 Application of RG 1.63 for Conforms The portion of the guidance that endorses 3.11.2 ronmental Qualification Environmental Design and IEEE 317-1983 is applicable. See RG 1.63 echanical and Electrical Qualification of Electrical Penetration entry in Table 1.9-2 with respect to the pment Assemblies other aspects of RG 1.63. S 3.11, Rev. 0: II.4 Application of RG 1.73 for Conforms None. 3.11.2 ronmental Qualification Environmental Design and echanical and Electrical Qualification of Class 1E Electric Valve pment Operators S 3.11, Rev. 0: II.5 Application of RG 1.89 for Partially Conforms See RG 1.89 in Table 1.9-2. 3.11.2 ronmental Qualification Environmental Qualification of echanical and Electrical Electrical Equipment Important to pment Safety S 3.11, Rev. 0: II.6 Application of RG 1.97 for Partially Conforms See RG 1.97 in Table 1.9-2. 3.11.2 ronmental Qualification Environmental Design and echanical and Electrical Qualification of PostAccident pment Monitoring Equipment S 3.11, Rev. 0: II.7 Application of RG 1.152 for Conforms None. 3.11 Conformance with Regulatory Criteria ronmental Qualification Environmental design and echanical and Electrical qualification of computer-specific pment requirements S 3.11, Rev. 0: II.8 Application of RG 1.153 for Conforms None. 3.11 ronmental Qualification Environmental design and echanical and Electrical qualification of power, pment instrumentation, and control portions of the safety systems

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 3.11, Rev. 0: II.9 Application of RG 1.209 for Conforms None. 3.11 ronmental Qualification Environmental design and echanical and Electrical qualification of safety-related pment computer-based I&C systems in mild environments S 3.11, Rev. 0: II.10 Application of RG 1.211 for Conforms None. 3.11.2 ronmental Qualification Environmental Qualification of Class echanical and Electrical 1E Electric Cables and Field Splices pment S 3.11, Rev. 0: II.11 Application of RG 1.156 for Conforms None. 3.11.2 ronmental Qualification Environmental Qualification of Class echanical and Electrical 1E Connection Assemblies pment S 3.11, Rev. 0: II.12 Application of RG 1.158 for Not Applicable See RG 1.158 in Table 1.9-2. Not Applicable ronmental Qualification Environmental Qualification of Class echanical and Electrical 1E Lead Storage Batteries pment S 3.11, Rev. 0: II.13 Application of RG 1.180 for Partially Conforms See RG 1.180 in Table 1.9-2. 3.11.2 ronmental Qualification Electromagnetic and Radio-echanical and Electrical Frequency Interference in Safety pment Related I&C Equipment S 3.11, Rev. 0: II.14 Application of RG 1.183 for Accident Partially Conforms See RG 1.183 in Table 1.9-2. 3.11.2 ronmental Qualification Source Term Used in Environmental echanical and Electrical Design and Qualification of pment Equipment Important to Safety Conformance with Regulatory Criteria S 3.11, Rev. 0: II.15 Application of RG 1.100 for Seismic Partially Conforms See RG 1.100 in Table 1.9-2. 3.11 ronmental Qualification Qualification of Electrical and Active echanical and Electrical Mechanical Equipment and pment Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants S 3.11, Rev. 0: II.16 Application of RG 1.204 for Not Applicable Lightning protection is not applicable to EQ Not Applicable ronmental Qualification Environmental design and because it is associated with an external/ echanical and Electrical qualification of the lightning natural event. pment protection system

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 3.11, Rev. 0: II.17 Effects of Environmental Conditions Conforms None. 3.11 ronmental Qualification for All Important to Safety Equipment echanical and Electrical pment S 3.11, Rev. 0: II.18 Suitability of Materials, Parts, and Conforms None. 3.11 ronmental Qualification Equipment Essential to Safety-echanical and Electrical Related Functions pment S 3.11, Rev. 0: II.19 Qualification of Nonmetallic Parts Conforms None. 3.11 ronmental Qualification echanical and Electrical pment S 3.11, Rev. 0: II.20 Design/Purchase Specifications of Conforms None. 3.11 ronmental Qualification Equipment to Perform Under echanical and Electrical Applicable Environmental Conditions pment S 3.11, Rev. 0: II.21 Applicable documentation for Conforms None. 3.11 ronmental Qualification Environmental Design and echanical and Electrical Qualification of Safety-Related pment Mechanical, Electrical, and I&C Equipment S 3.11, Rev. 0: II.22 Maintenance/surveillance programs Not Applicable The programs are described and Not Applicable ronmental Qualification to provide assurance Assurance of maintained by the COL applicant. echanical and Electrical Environmental Design and pment Qualification Status of Equipment in Conformance with Regulatory Criteria Mild and Harsh Environments S 3.11, Rev. 0: II.23 Operational Program Not Applicable This is a COL applicant item. Not Applicable ronmental Qualification Implementation echanical and Electrical pment S 3.11, Rev. 0: II.24 Exposure of Organic Components on Conforms None. 3.11 ronmental Qualification Engineered Safety Features Systems echanical and Electrical pment

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 3.11, Rev. 0: II.25 Design and Procurement Not Applicable This is a COL applicant item. Not Applicable ronmental Qualification Specifications echanical and Electrical pment 3.12, Rev 1: ASME Code II.A Piping Analysis Methods Conforms None. 3.12.3 1, 2, and 3 Piping ems, Piping Components Their Associated orts 3.12, Rev 1: ASME Code II.B Piping Modeling Techniques Conforms None. 3.12.4 1, 2, and 3 Piping ems, Piping Components Their Associated orts 3.12, Rev 1: ASME Code II.C Piping Stress Analysis Criteria Conforms None. 3.12.5 1, 2, and 3 Piping ems, Piping Components Their Associated orts 3.12, Rev 1: ASME Code II.D Piping Support Design Conforms None. 3.12.6 1, 2, and 3 Piping ems, Piping Components Their Associated orts 3.13, Rev. 0: Threaded II.1 Design Aspects (Including Table 3.13- Conforms None. 3.13.1 Conformance with Regulatory Criteria eners - ASME Code Class 1) and 3 3.13, Rev. 0: Threaded II.2 Preservice and Inservice Inspection Conforms None. 3.13.2 eners - ASME Code Class Requirements (Including Table 3.13-and 3 2) 3-1, Rev 2: Classification All Not Applicable This guidance is applicable only to BWR Not Applicable ain Steam Components plants. r Than the Reactor ant Pressure Boundary WR Plants

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 3-2, Rev 2: Classification All Not Applicable This guidance is applicable only to BWR Not Applicable WR/6 Main Steam and plants. water Components r Than the Reactor ant Pressure Boundary 3-3, Rev 3: Protection B.1 Plant Arrangement Conforms None. 3.6 nst Postulated Piping res in Fluid Systems ide Containment 3-3, Rev 3: Protection B.2 Design Features Conforms None. 3.6 nst Postulated Piping res in Fluid Systems ide Containment 3-3, Rev 3: Protection B.3 Analyses and Effects of Postulated Conforms None. 3.6 nst Postulated Piping Piping Failures res in Fluid Systems ide Containment 3-3, Rev 3: Protection B.4 Implementation Conforms None. 3.6 nst Postulated Piping res in Fluid Systems ide Containment 3-4, Rev 2: Postulated B.A High-Energy Fluid System Piping Conforms None. 3.6 ure Locations in Fluid 15.1 em Piping Inside and 15.2 ide Containment 15.5 Conformance with Regulatory Criteria 15.6 3-4, Rev 2: Postulated B.B Moderate-Energy Fluid System Piping Conforms None. 3.6 ure Locations in Fluid 15.1 em Piping Inside and 15.2 ide Containment 15.5 15.6 3-4, Rev 2: Postulated B.C Type of Breaks and Leakage Cracks in Conforms None. 3.6 ure Locations in Fluid Fluid System Piping 15.1 em Piping Inside and 15.2 ide Containment 15.5 15.6

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 4.2, Rev 3: Fuel System II.1 Design Bases Conforms None. 4.2.1 gn 4.2, Rev 3: Fuel System II.1.A Fuel System Damage Conforms None. 4.2.1 gn 4.2, Rev 3: Fuel System II.1.B Fuel Rod Failure Conforms None. 4.2.1 gn 4.2.3 4.2, Rev 3: Fuel System II.1.C Fuel Coolability Conforms None. 4.2.1 gn 4.2, Rev 3: Fuel System II.2 Description and Design Drawings Conforms None. 4.2.2 gn 4.2, Rev 3: Fuel System II.3 Design Evaluation Conforms None. 4.2.1 gn 4.2.3 4.2.4 4.2, Rev 3: Fuel System II.4 Testing, Inspection, and Surveillance Conforms None. 4.2.1 gn Plans 4.2.4 4.2, Rev 3: Fuel System App A Evaluation of Fuel Assembly Conforms None. 4.2.1 gn Structural Response to Externally Applied Forces 4.2, Rev 3: Fuel System App B Interim Acceptance Criteria and Conforms None. 4.2.1 gn Guidance for the Reactivity Initiated 15.0.0 Accidents 4.3, Rev 0: II.1 Design Limits for Power Densities and Conforms None. 4.3.1 ear Design Power Distributions 4.3, Rev 0: II.2 Reactivity Coefficients Conforms None. 4.3.2 Conformance with Regulatory Criteria ear Design 4.3, Rev 0: II.3 Control Rod Patterns and Reactivity Conforms None. 4.3.2 ear Design Worth 4.3, Rev 0: II.4 Analytical Methods and Data Conforms None. 4.3.3 ear Design S 4.4, Rev 0: Thermal and II.1 Fuel Design Limits, Core Design, and Conforms None. 4.4.1 aulic Design Thermal Margin 4.4.2 S 4.4, Rev 0: Thermal and II.2 Subchannel Hydraulic Analysis Codes Conforms None. 4.4.4 aulic Design S 4.4, Rev 0: Thermal and II.3 Core Oscillations and Thermal- Conforms None. 4.4.7 aulic Design Hydraulic Instabilities

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 4.4, Rev 0: Thermal and II.4 RPV Fluid Flow Calculations Conforms None. 4.4.4 aulic Design S 4.4, Rev 0: Thermal and II.5 Technical Specifications Conforms None. 4.4.3 aulic Design 4.4.6 16.1 S 4.4, Rev 0: Thermal and II.6 Preoperational and Initial Test Conforms None. 4.4.5 aulic Design Programs S 4.4, Rev 0: Thermal and II.7 Loose Parts Monitoring System Departure Low flow in primary systems precludes 4.4.6 aulic Design damage from loose parts and the need for loose parts monitoring system. S 4.4, Rev 0: Thermal and II.8 Critical Heat Flux Calculations and Conforms None. 4.4.2 aulic Design Process Monitoring 4.4.4 4.4.6 S 4.4, Rev 0: Thermal and II.9 Instrumentation and Procedures for Conforms None. 4.4.6 aulic Design Detection and Recovery from Inadequate Core Cooling S 4.4, Rev 0: Thermal and II.10 Core Stability Performance During Not Applicable Diverse RTS signals prevent an ATWS from Not Applicable aulic Design Anticipated Transient without Scram occurring. This prevents flow instabilities Event from occurring, so this AC is not applicable based on the current ATWS approach. 4.5.1, Rev 3: Control Rod II.1 Materials Specifications Conforms RG 1.85 was withdrawn in 2004. Guidance 4.5.1 e Structural Materials was updated and incorporated into RG 1.84. 4.5.1, Rev 3: Control Rod II.2 Austenitic Stainless Steel Conforms The NuScale QAPD is based on ANSI/ASME 4.5.1 e Structural Materials Components NQA-1-2008 with NQA-1a-2009 addenda, as Conformance with Regulatory Criteria endorsed by RG 1.28, Rev. 4. 4.5.1, Rev 3: Control Rod II.3 Other Materials Conforms None. 4.5.1 e Structural Materials 4.5.1, Rev 3: Control Rod II.4 Cleaning and Cleanliness Control Conforms The NuScale QAPD is based on ANSI/ASME 4.5.1 e Structural Materials NQA-1-2008 with NQA-1a-2009 addenda, as endorsed by RG 1.28, Rev. 4. 4.5.2, Rev 3: Reactor II.1 Materials Conforms None. 4.5.2 nal and Core Support cture Materials 4.5.2, Rev 3: Reactor II.2 Controls on Welding Conforms None. 4.5.2 nal and Core Support cture Materials

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 4.5.2, Rev 3: Reactor II.3 Nondestructive Examination Conforms None. 4.5.2 nal and Core Support cture Materials 4.5.2, Rev 3:Reactor II.4 Austenitic Stainless Steels Conforms None. 4.5.2 nal and Core Support cture Materials 4.5.2, Rev 3: Reactor II.5 Other Materials Conforms None. 4.5.2 nal and Core Support cture Materials 4.6, Rev 2: Functional II.1 Environmental and Dynamic Effects - Conforms None. 4.6.2 gn of Control Rod Drive GDC 4 em 4.6, Rev 2: Functional II.2 Failure Modes and Effects - GDC 23 Conforms None. 4.6.2 gn of Control Rod Drive em 4.6, Rev 2: Functional II.3 Single Malfunction - GDC 25 Conforms None. 4.6.2 gn of Control Rod Drive em 4.6, Rev 2: Functional II.4 Operational Control and Reliability - Conforms NuScale does not interpret GDC 26 as 4.6.2 gn of Control Rod Drive GDC 26 requiring two safety-related means of em reactivity control. One of the independent reactivity control systems used to meet the requirements of GDC 26 in the NuScale design is the chemical and volume control system, which is not safety-related. Conformance with Regulatory Criteria 4.6, Rev 2: Functional II.5 Combined Capability - GDC 27 Departure The NuScale design bases conform to a 3.1 gn of Control Rod Drive design-specific Principal Design Criterion 4.2 em (PDC) in lieu of GDC 27, as reflected in 4.3 Section 3.1. 4.6.2 4.6, Rev 2: Functional II.6 Reactivity Limits - GDC 28 Conforms None. 4.6.0.2 gn of Control Rod Drive 4.6.2 em 4.6, Rev 2: Functional II.7 Protection Against Anticipated Conforms None. 4.6.2 gn of Control Rod Drive Operational Occurrences - GDC 29 em

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 4.6, Rev 2: Functional II.8 BWR Alternate Rod Injection System Not Applicable This guidance is applicable only to BWR Not Applicable gn of Control Rod Drive plants. em 4-1, Rev 3: Westinghouse All - Not Applicable This BTP is applicable only to PWR designs Not Applicable tant Axial Offset Control that use the Constant Axial Offset Control C) operating scheme. NuScale does not use the Constant Axial Offset Control operating scheme. 5.2.1.1, Rev 3: II Use of RG 1.26 to meet GDC 1 and Conforms See RG 1.26 in Table 1.9-2. 5.2.1 pliance with the Codes 10 CFR 50.55a Standards Rule, FR 50.55a 5.2.1.2, Rev 3: Applicable II.1 Use of RG 1.84 to meet GDC 1 and Conforms See RG 1.26 in Table 1.9-2. 5.2.1 e Cases 10 CFR 50.55a 5.2.1.2, Rev 3: Applicable II.2 Use of RG 1.147 to meet GDC 1 and Partially Conforms See RG 1.147 in Table 1.9-2. 5.2.1 e Cases 10 CFR 50.55a 5.2.1.2, Rev 3: Applicable II.3 Use of RG 1.192 to meet GDC 1 and Partially Conforms See RG 1.192 in Table 1.9-2. 5.2.1 e Cases 10 CFR 50.55a 5.2.2, Rev 3: Overpressure II.1 Material Specifications Conforms None. 5.2.2 ection 5.2.2, Rev 3: Overpressure II.2 Design Requirements for BWRs Not Applicable This guidance is applicable only to BWR Not Applicable ection Operating at Power plants. 5.2.2, Rev 3: Overpressure II.3 Design Requirements for PWRs Partially Conforms The overpressure analysis does not assume 5.2.2 ection Operating at Power a secondary safety-grade signal from the Conformance with Regulatory Criteria RPS initiates the reactor trip. NuScale does not have a secondary safety-grade reactor trip system. 5.2.2, Rev 3: Overpressure II.4 Design Requirements for PWRs Conforms None. 5.2.2 ection Operating at Low Temperature (Startup, Shutdown) 5.2.2, Rev 3: Overpressure II.5 Testing and Inspections Conforms None. 5.2.2 ection 5.2.2, Rev 3: Overpressure II.6 Technical Specifications Partially Conforms Certain subtier guidance documents 5.2.2 ection referenced in this acceptance criterion are not applicable or only partially applicable.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 5.2.2, Rev 3: Overpressure II.7 TMI Action Plan Requirements Conforms None. 5.2.2 ection 5.2.3, Rev 3: Reactor II.1 Material Specifications Partially Conforms This acceptance criterion is applicable 5.2.3 ant Pressure Boundary except for references to subtier guidance rials that is applicable only to BWRs. 5.2.3, Rev 3: Reactor II.2 Compatibility of Materials with the Partially Conforms This acceptance criterion is applicable 5.2.3 ant Pressure Boundary Reactor Coolant except for references to subtier guidance rials that is applicable only to BWRs. 5.2.3, Rev 3: Reactor II.3 Fabrication and Processing of Ferritic Conforms None. 5.2.3 ant Pressure Boundary Materials rials 5.2.3, Rev 3: Reactor II.3.A Fracture Toughness - 10 CFR 50, Conforms None. 5.2.3 ant Pressure Boundary Appendix G rials 5.2.3, Rev 3: Reactor II.3.B Control of Ferritic Steel Welding Conforms None. 5.2.3 ant Pressure Boundary rials 5.2.3, Rev 3: Reactor II.3.C NDE of Ferritic Steel Tubular Products Conforms None. 5.2.3 ant Pressure Boundary rials 5.2.3, Rev 3: Reactor II.4 Fabrication and Processing of Partially Conforms This acceptance criterion is applicable 5.2.3 ant Pressure Boundary Austenitic Stainless Steel except for references to subtier guidance rials that is applicable only to BWRs or to large LWRs that use nonmetallic thermal Conformance with Regulatory Criteria insulation on reactor coolant pressure boundary components. 5.2.3, Rev 3: Reactor II.4.A GDC 4 Compatibility of Components - Partially Conforms This acceptance criterion is applicable 5.2.3 ant Pressure Boundary Measures to Avoid Sensitization in except for references to subtier guidance rials Austenitic Stainless Steel that is applicable only to BWRs. 5.2.3, Rev 3: Reactor II.4.B GDC 4 Compatibility of Components - Partially Conforms This acceptance criterion is applicable 5.2.3 ant Pressure Boundary Controls to Avoid Stress Corrosion except for references to subtier guidance rials Cracking in Austenitic Stainless Steel that is applicable only to BWRs, and to subtier RG 1.37, which endorses use of NQA-1-1994. The NuScale design is based on NQA-1-2008 and the NQA-1a-2009 addenda, rather than NQA-1-1994.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 5.2.3, Rev 3: Reactor II.4.C Compatibility of Austenitic Stainless Not Applicable This acceptance criterion is applicable only Not Applicable ant Pressure Boundary Steel Materials with Thermal to LWRs that use nonmetallic thermal rials Insulation insulation on reactor coolant pressure boundary components. NuScale does not use nonmetallic thermal insulation on reactor coolant pressure boundary components. 5.2.3, Rev 3: Reactor II.4.D Control of Welding of Austenitic Partially Conforms This acceptance criterion is applicable 5.2.3 ant Pressure Boundary Stainless Steels except for references to subtier guidance rials that is applicable only to BWRs. 5.2.3, Rev 3: Reactor II.4.E NDE of Austenitic Stainless Steel Conforms None. 5.2.3 ant Pressure Boundary Tubular Products rials 5.2.3, Rev 3: Reactor II.4.G Operational Programs Not Applicable This acceptance criterion governs plant- Not Applicable ant Pressure Boundary specific programmatic information that is rials the responsibility of the COL applicant. S 5.2.4, Rev 0: Reactor II.1 System Boundary Subject to Conforms None. 5.2.4 ant Pressure Boundary Inspection vice Inspection and ng S 5.2.4, Rev 0: Reactor II.2 Accessibility Conforms None. 5.2.4 ant Pressure Boundary vice Inspection and ng S 5.2.4, Rev 0: Reactor II.3 Examination Categories and Methods Conforms None. 5.2.4 Conformance with Regulatory Criteria ant Pressure Boundary vice Inspection and ng S 5.2.4, Rev 0: Reactor II.4 Inspection Intervals Conforms None. 5.2.4 ant Pressure Boundary vice Inspection and ng S 5.2.4, Rev 0: Reactor II.5 Evaluation of Examination Results Conforms None. 5.2.4 ant Pressure Boundary vice Inspection and ng

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 5.2.4, Rev 0: Reactor II.6 System Pressure Tests Conforms None. 5.2.4 ant Pressure Boundary vice Inspection and ng S 5.2.4, Rev 0: Reactor II.7 Code Exemptions Conforms None. 5.2.4 ant Pressure Boundary vice Inspection and ng S 5.2.4, Rev 0: Reactor II.8 Relief Requests Conforms None. 5.2.4 ant Pressure Boundary vice Inspection and ng S 5.2.4, Rev 0: Reactor II.9 Code Cases Conforms None. 5.2.4 ant Pressure Boundary vice Inspection and ng S 5.2.4, Rev 0: Reactor II.10 Augmented ISI to Protect Against Conforms None. 5.2.4 ant Pressure Boundary Postulated Piping Failures vice Inspection and ng S 5.2.4, Rev 0: Reactor II.11 Other Inspection Programs Partially Conforms Although a boric acid control program has 5.2.4 ant Pressure Boundary not been established, a brief description of vice Inspection and the program is provided in the DCA. ng S 5.2.4, Rev 0: Reactor II.12 Operational Programs Not Applicable This acceptance criterion governs plant- Not Applicable Conformance with Regulatory Criteria ant Pressure Boundary specific programmatic information that is vice Inspection and the responsibility of the COL applicant. ng S 5.2.4, Rev 0: Reactor II.13 ITAAC Partially Conforms A portion of this acceptance criterion is 5.2.4 ant Pressure Boundary applicable only to COL applicants. vice Inspection and ng S 5.2.4, Rev 0: Reactor II.14 Risk Informed ISI Program Not Applicable This acceptance criterion governs plant- Not Applicable ant Pressure Boundary specific programmatic information that is vice Inspection and the responsibility of the COL applicant. ng

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 5.2.5, Rev 0: Reactor II.1 Criteria to Meet GDC 2 Conforms None. 5.2.5 ant Pressure Boundary age Detection S 5.2.5, Rev 0: Reactor II.2 Criteria to Meet GDC 14 Conforms None. 5.2.5 ant Pressure Boundary age Detection S 5.2.5, Rev 0: Reactor II.3 Criteria to Meet GDC 30 Conforms None. 5.2.5 ant Pressure Boundary age Detection S 5.3.1, Rev 0: Reactor II.1 Materials Conforms None. 5.3.1 el Materials S 5.3.1, Rev 0: Reactor II.2 Special Processes Used for Conforms None. 5.3.1 el Materials Manufacture and Fabrication of Components S 5.3.1, Rev 0: Reactor II.3 Special Methods for Nondestructive Conforms None. 5.3.1 el Materials Examination S 5.3.1, Rev 0: Reactor II.4 Special Controls and Special Conforms None. 5.3.1 el Materials Processes Used for Ferritic Steels and Austenitic Stainless Steels S 5.3.1, Rev 0: Reactor II.5 Fracture Toughness Conforms None. 5.3.1 el Materials S 5.3.1, Rev 0: Reactor II.6 Material Surveillance Conforms None. 5.3.1 el Materials S 5.3.1, Rev 0: Reactor II.7 Reactor Vessel Fasteners Conforms None. 5.3.1 Conformance with Regulatory Criteria el Materials S 5.3.2, Rev 0: Pressure- II.1.A Pressure-Temperature - Applicable Conforms None. 5.3.2 perature Limits, Upper- Regulations, Codes, and Basis f Energy, and Pressurized Documents mal Shock S 5.3.2, Rev 0: Pressure- II.1.B Pressure-Temperature Requirements Conforms None. 5.3.2 perature Limits, Upper-f Energy, and Pressurized mal Shock

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 5.3.2, Rev 0: Pressure- II.2.A Upper-Shelf Energy - Applicable Conforms None. 5.3.2 perature Limits, Upper- Regulations, Codes, and Basis f Energy, and Pressurized Documents mal Shock S 5.3.2, Rev 0: Pressure- II.2.B Upper-Shelf Energy Requirements Conforms None. 5.3.1 perature Limits, Upper- 5.3.2 f Energy, and Pressurized mal Shock S 5.3.2, Rev 0: Pressure- II.3.A Pressurized Thermal Shock - Conforms None. 5.3.2 perature Limits, Upper- Applicable Regulations, Codes, and f Energy, and Pressurized Basis Documents mal Shock S 5.3.2, Rev 0: Pressure- II.3.B Pressurized Thermal Shock Conforms None. 5.3.2 perature Limits, Upper- Requirements f Energy, and Pressurized mal Shock S 5.3.3, Rev 0: Reactor II.1 Design Conforms None. 5.3.3 el Integrity S 5.3.3, Rev 0: Reactor II.2 Materials of Construction Conforms None. 5.3.3 el Integrity S 5.3.3, Rev 0: Reactor II.3 Fabrication Methods Conforms None. 5.3.3 el Integrity S 5.3.3, Rev 0: Reactor II.4 Inspection Requirements Conforms None. 5.3.3 el Integrity Conformance with Regulatory Criteria S 5.3.3, Rev 0: Reactor II.5 Shipment and Installation Conforms None. 5.3.3 el Integrity S 5.3.3, Rev 0: Reactor II.6 Operating Conditions Conforms None. 5.3.3 el Integrity S 5.3.3, Rev 0: Reactor II.7 Inservice Surveillance Conforms Inservice surveillance of the reactor vessel is 5.3.3 el Integrity described in the DCD. However, the COL applicant develops and implements the reactor vessel surveillance program. S 5.3.3, Rev 0: Reactor II.8 Operational Programs Not Applicable This acceptance criterion governs plant- Not Applicable el Integrity specific programmatic information that is the responsibility of the COL applicant.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 5.3.3, Rev 0: Reactor II.9 10 CFR 52.47(b)(1) compliance Not Applicable This requirement applies to plant-specific Not Applicable el Integrity verification and is the responsibility of the COL applicant. 5.4.1.1, Rev 3: Pump All Various Not Applicable This SRP section and its acceptance criteria Not Applicable heel Integrity (PWR) (II.1 through II.7) apply only to PWR designs that use reactor coolant pumps. The NuScale reactor design does not require or include reactor coolant pumps. Rather, the NuScale design uses passive natural circulation of the primary coolant, eliminating the need for reactor coolant pumps. S 5.4.2.1, Rev 0: Steam II.1 Selection, Processing, Testing, and Conforms None. 5.4.1 erator Materials Inspection of Materials S 5.4.2.1, Rev 0: Steam II.2 Steam Generator Design Conforms None. 5.4.1 erator Materials S 5.4.2.1, Rev 0: Steam II.3 Fabrication and Processing of Ferritic Conforms None. 5.4.1 erator Materials Materials S 5.4.2.1, Rev 0: Steam II.4 Fabrication and Processing of Conforms None. 5.2 erator Materials Austenitic Stainless Steel 5.4.1 S 5.4.2.1, Rev 0: Steam II.5 Compatibility of Materials with the Conforms None. 5.4.1 erator Materials Primary (Reactor) and Secondary Coolant and Cleanliness Control S 5.4.2.1, Rev 0: Steam II.6 Provisions for Accessing the Conforms None. 5.4.1 Conformance with Regulatory Criteria erator Materials Secondary Side of the Steam Generator S 5.4.2.2, Rev 0: Steam II.1 Steam Generator Tube Susceptibility Conforms None. 5.4 erator Program to Degradation S 5.4.2.2, Rev 0: Steam II.2 Steam Generator Monitoring Partially Conforms A portion of this acceptance criterion is 5.4.1 erator Program Program Elements applicable to COL applicants referencing a certified design. S 5.4.2.2, Rev 0: Steam II.3 Steam Generator Program Elements Partially Conforms Certain subtier guidance documents 5.4.1 erator Program in Technical Specifications referenced in this acceptance criterion are only partially applicable. S 5.4.2.2, Rev 0: Steam II.4 Steam Generator Tube Repair Criteria Conforms None. 5.4.1 erator Program

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 5.4.2.2, Rev 0: Steam II.5 Steam Generator Tube Repair Conforms None. 5.4.1 erator Program Methods S 5.4.2.2, Rev 0: Steam II.6 Steam Generator Tube Preservice Conforms None. 5.4.1 erator Program Inspection S 5.4.2.2, Rev 0: Steam II.7 Periodic Tube Inspection and Testing Partially Conforms 5.4.1 erator Program in Certified Design Technical Specifications S 5.4.2.2, Rev 0: Steam II.8 Operational Programs Partially Conforms This acceptance criterion governs plant- 5.4.1 erator Program specific programmatic activities that are the responsibility of the COL applicant referencing a certified design. S 5.4.2.2, Rev 0: Steam II.9 ITAAC Partially Conforms A portion of this acceptance criterion is 5.4.1 erator Program applicable only to COL applicants. 5.4.6, Rev 4: Reactor Core All Various Not Applicable This SRP section and its acceptance criteria Not Applicable tion Cooling System (II.1 through II.10) apply only to BWRs. R) S 5.4.7, Rev 0: Decay Heat II.1 thru II.3 Various Conforms None. 5.4.3 oval (DHR) System onsibilities S 5.4.7, Rev 0: Decay Heat II.4 GDC 5 Conforms None. 5.4.3 oval (DHR) System onsibilities S 5.4.7, Rev 0: Decay Heat II.5 GDC 14 Not Applicable The DHRS is connected to the secondary Not Applicable oval (DHR) System system and does not directly interface with Conformance with Regulatory Criteria onsibilities the RCPB. S 5.4.7, Rev 0: Decay Heat II.6 GDC 19 Departure The NuScale design supports an exemption 5.4.3 oval (DHR) System from GDC 19. As described in Section 3.1.2, onsibilities the design complies with a NuScale-specific principal design criterion (PDC) in lieu of this GDC. S 5.4.7, Rev 0: Decay Heat II.7 GDC 34 Departure The NuScale design supports an exemption 5.4.3 oval (DHR) System from the power provisions of GDC 34. As onsibilities described in Section 3.1.4, the design complies with a NuScale-specific principal design criterion in lieu of this GDC.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 5.4.7, Rev 0: Decay Heat II.8 GDC 54 Partially Conforms This closed-loop DHRS outside the 5.4.3 oval (DHR) System containment is directly connected to the onsibilities closed-loop SG system within the RPV providing dual passive barriers between the RCS and the reactor pool outside the NPM. Breaches of this piping system outside containment is not considered credible because the system is a welded design with a system design pressure equivalent to the RPV, designed to Class 2 requirements in accordance with ASME BPV Code, Section III, and meets the applicable criteria of NRC Branch Technical Position 3-4, Revision 2. As a result, leakage detection and isolation capabilities of this piping system from containment are not considered important to safety. S 5.4.7, Rev 0: Decay Heat II.9 DHRS Interface with other systems Conforms None. 5.4.3 oval (DHR) System onsibilities 5.4.8, Rev 3: Reactor All Various Not Applicable This SRP section and its acceptance criteria Not Applicable er Cleanup System (BWR) (II.1 through II.4) apply only to BWRs. 5.4.11, Rev 4: Pressurizer All Various Not Applicable This SRP section and its acceptance criteria Not Applicable f Tank (II.1 and II.2) apply only to PWRs that use a pressurizer relief tank. A pressurizer relief Conformance with Regulatory Criteria tank is not used in the NuScale design. Fluid relieved through the reactor coolant system overpressure protection system is routed directly to the containment vessel.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 5.4.12, Rev 1: Reactor All Various Departure Because of the integral reactor coolant Not Applicable ant System High Point system configuration, non-condensable s gases accumulating in the pressurizer space will not interfere with core cooling during or after design basis accidents. The NuScale design supports an exemption from the requirements of 10 CFR 50.46a related to reactor coolant system high point venting, as well as the substantively similar requirements of 10 CFR 50.34(f)(2)(vi). 5.4.13, (March 2007): All Various Not Applicable This SRP section and its acceptance criteria Not Applicable tion Condenser System (II.1 through II.12) are applicable only to R) BWRs. 5-1, Rev 3: Monitoring of B.1 Secondary Water Chemistry Program Conforms None. 5.4.1 ndary Side Water Meeting Industry Guidelines 10.3.5 mistry in PWR Steam erators 5-1, Rev 3: Monitoring of B.2 Sampling Schedule for Critical Partially Conforms A portion of this acceptance criterion 5.4.1 ndary Side Water Parameters governs information that is site-specific and 10.3.5 mistry in PWR Steam is the responsibility of the COL applicant erators referencing the certified design. 5-1, Rev 3: Monitoring of B.3 Records Partially Conforms A portion of this acceptance criterion 5.4.1 ndary Side Water governs information that is site-specific and 10.3.5 mistry in PWR Steam is the responsibility of the COL applicant erators referencing the certified design. Conformance with Regulatory Criteria 5-1, Rev 3: Monitoring of B.4 Program Change Evaluation and Not Applicable This acceptance criterion is the Not Applicable ndary Side Water Reporting responsibility of the COL applicant mistry in PWR Steam referencing the certified design. erators 5-2, Rev 3: B.1 System Design, Installation, and Conforms None. 5.2.2 pressurization Capabilities to Prevent Exceeding ection of Pressurized- Technical Specifications and NRC er Reactors While Regulatory Requirements rating at Low peratures

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 5-2, Rev 3: B.2 Low-Temperature Overpressure Partially Conforms Conforms to ASME Section XI Appendix G 5.2.2 pressurization Protection Operability Criteria. ection of Pressurized-er Reactors While rating at Low peratures 5-2, Rev 3: B.3 System Designed to Withstand Single Conforms None. 5.2.2 pressurization Active Component Failure ection of Pressurized-er Reactors While rating at Low peratures 5-2, Rev 3: B.4 System Instrumentation and Controls Conforms None. 5.2.2 pressurization Design ection of Pressurized-er Reactors While rating at Low peratures 5-2, Rev 3: B.5 System Operability Testing Conforms None. 5.2.2 pressurization Ch 16 ection of Pressurized-er Reactors While rating at Low peratures 5-2, Rev 3: B.6 Applicable Guidance Conforms None. 5.2.2 Conformance with Regulatory Criteria pressurization ection of Pressurized-er Reactors While rating at Low peratures 5-2, Rev 3: B.7 System Design to Withstand Conforms None. 5.2.2 pressurization Operating-Basis Earthquake ection of Pressurized-er Reactors While rating at Low peratures

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 5-2, Rev 3: B.8 Backup Electrical Power Source Partially Conforms The intent of this guidance - that the low 5.2.2 pressurization temperature overpressure protection ection of Pressurized- (LTOP) system should not depend on the er Reactors While availability of offsite power to perform its rating at Low function - applies to the NuScale design. peratures 5-2, Rev 3: B.9 Analyses Considering Inadvertent Conforms None. 5.2.2 pressurization System Actuation ection of Pressurized-er Reactors While rating at Low peratures 5-2, Rev 3: B.10 Interlocks to Ensure Overpressure Partially Conforms The intent of this acceptance criterion is 5.2.2 pressurization Protection applicable but the criterion refers to large ection of Pressurized- LWR designs that provide pressure relief er Reactors While from a low-pressure system not normally rating at Low connected to the primary system. In the peratures NuScale design, the LTOP system is not connected to a low-pressure system. However, the intent of this guidance - to ensure that the LTOP system is not inadvertently isolated from the primary system - is applicable to the DCA. 5-3, Rev 2: Fracture 1 Preservice Fracture Toughness Test Partially Conforms This acceptance criterion is applicable 5.3 hness Requirements Requirements except as indicated in the comments below Conformance with Regulatory Criteria for Acceptance Criteria 1.1 and 1.2. 5-3, Rev 2: Fracture 1.1 Determination of RTNDT for Vessel Partially Conforms Portions of this acceptance criterion apply 5.3 hness Requirements Materials only to older plants for which fracture toughness testing on vessel material did not include all tests necessary to determine RTNDT. The rest of this guidance applies to the NuScale design. 5-3, Rev 2: Fracture 1.2 Estimation of Charpy V-Notch Upper Partially Conforms This guidance is applicable except for 5.3 hness Requirements Shelf Energies reference to subtier NUREG-0744, which applies only to operating reactors that do not meet the minimum fracture toughness acceptance criteria defined in this BTP 5-3.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 5-3, Rev 2: Fracture 1.3 Reporting Requirements Conforms None. 5.3 hness Requirements 5-3, Rev 2: Fracture 2 Operating Limitations for Fracture Conforms None. 5.3 hness Requirements Toughness 5-3, Rev 2: Fracture 2.1 Pressure-Temperature Operating Conforms None. 5.3 hness Requirements Limitations 5-3, Rev 2: Fracture 2.2 Recommended Bases for Operating Conforms None. 5.3 hness Requirements Limitations 5-3, Rev 2: Fracture 2.3 Reporting Requirements Conforms None. 5.3 hness Requirements 5-3, Rev 2: Fracture 3 Inservice Surveillance of Fracture Partially Conforms This acceptance criterion applies except as 5.3 hness Requirements Toughness indicated in the comments below for Acceptance Criteria 3.4 and 3.5. 5-3, Rev 2: Fracture 3.1 Surveillance Program Requirements Conforms None. 5.3 hness Requirements 5-3, Rev 2: Fracture 3.2 SAR Criteria Conforms None. 5.3 hness Requirements 5-3, Rev 2: Fracture 3.3 Surveillance Test Procedures Conforms None. 5.3 hness Requirements 5-3, Rev 2: Fracture 3.4 Reporting Criteria Not Applicable This acceptance criterion governs plant- Not Applicable hness Requirements specific reporting criteria that are the responsibility of the COL applicant that references the NuScale certified design. 5-3, Rev 2: Fracture 3.5 Technical Specification Changes Not Applicable This acceptance criterion governs plant- Not Applicable Conformance with Regulatory Criteria hness Requirements specific activities that are the responsibility of the COL applicant that references the NuScale certified design. 5-3, Rev 2: Fracture 4.1 Pressurized Thermal Shock Conforms None. 5.3 hness Requirements Requirements S BTP 5-4, Rev 0: Design B.1 Functional Requirements Conforms None. 5.4.3 irements of the Residual Removal System S BTP 5-4, Rev 0: Design B.2 Pressure Relief Requirements Conforms None. 5.4.3 irements of the Residual Removal System

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S BTP 5-4, Rev 0: Design B.3 Test Requirements Conforms None. 5.4.3 irements of the Residual Removal System S BTP 5-4, Rev 0: Design B.4 Operational Procedures Partially Conforms The procedures governed by this 5.4.3 irements of the Residual acceptance criterion are site-specific and Removal System are the responsibility of the COL applicant. S BTP 5-4, Rev 0: Design B.5 Implementation Conforms None. 5.4.3 irements of the Residual Removal System 6.1.1, Rev 2: Engineered II.1 Materials and Fabrication Conforms None. 6.1.1 ty Features Materials 6.1.1, Rev 2: Engineered II.1.A Austenitic Stainless Steels Conforms None. 6.1.1 ty Features Materials 6.1.1, Rev 2: Engineered II.1.B Ferritic Steel Welding Conforms None. 6.1.1 ty Features Materials 6.1.1, Rev 2: Engineered II.2 Composition and Compatibility of Conforms This guidance is applicable except the 6.1.1 ty Features Materials ESF Systems Fluids NuScale design does not provide a method for post-accident pH control as addressed in BTP 6-1. 6.1.1, Rev 2: Engineered II.3 Component and Systems Cleaning Partially Conforms RG 1.37 has been withdrawn by the NRC. 6.1.1 ty Features Materials 6.1.1, Rev 2: Engineered II.4 Thermal Insulation Conforms None. 6.1.1 ty Features Materials 6.1.2, Rev 3: Protective All Various Not Applicable This SRP section is applicable only to the Not Applicable Conformance with Regulatory Criteria ing Systems (Paints) - use of protective coatings on surfaces inside nic Materials the containment. The NuScale Power Module design does not use protective coatings inside the containment vessel. S 6.2.1, Rev 0: No specific Applicable acceptance criteria are See the applicable See SRP 2.4.6, 2.4.10, 2.4.12, 3.9.3, 19.0, and 2.4 ainment Functional requirements addressed in SRP 2.4.6, 2.4.10, 2.4.12, SRP or DSRS. DSRS 3.8.2. 3.8.2 gn listed. 3.9.3, 19.0 and DSRS 3.8.2. 3.9.3 6.2.1 19.2 S 6.2.1.1.A, Rev 0: II.1 Design Margin for Containment Conform The peak containment pressure for the 6.2.1 ainment Design Pressure limiting event is less than the design pressure.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 6.2.1.1.A, Rev 0: II.2 Reducing Containment Pressure Conforms None. 6.2.1 ainment Following Postulated Design Basis Accident S 6.2.1.1.A, Rev 0: II.3 Containment Heat Removal Conforms None. 6.2.1 ainment Capability and Design Margin - LOCA 6.2.2 Assumptions S 6.2.1.1.A, Rev 0: II.4 Containment Heat Removal Conforms None. 6.2.1 ainment Capability and Design Margin - 6.2.2 Containment Response Analysis Assumptions S 6.2.1.1.A, Rev 0: II.5 Protection of Containment from Conforms None. 6.2.1 ainment External Pressure Conditions S 6.2.1.1.A, Rev 0: II.6 Containment Monitoring Conforms None. 6.2.1 ainment Instrumentation S 6.2.1.1.A, Rev 0: II.7 Design of Containment Internal Conforms None. 6.2.1 ainment Structures and System Components S 6.2.1.1.A, Rev 0: II.8 Evaluation of Accident Involving Conforms None. 6.2.1 ainment Generated Hydrogen S 6.2.1.1.A, Rev 0: II.9 Evaluation of an Accident on other Conforms None. 6.2.1 ainment Modules 6.2.1.1.B, Draft Rev 3: Ice All Various Not Applicable The NuScale design does not use an ice Not Applicable denser Containments condenser containment. 6.2.1.1.C, Rev 7: Pressure All Various Not Applicable This SRP section and its acceptance criteria Not Applicable pression Type BWR apply only to applicants for BWR designs Conformance with Regulatory Criteria ainments that involve Pressure Suppression Type Containments. 6.2.1.2, Rev 3: All Various Not Applicable This SRP section and its acceptance criteria Not Applicable ompartment Analysis (II.1 through II.4) are applicable only to LWR designs that involve a containment structure that houses subcompartments. The NuScale containment vessel design does not have subcompartments housing high-energy piping as defined in this guidance (or internal compartments as referred to in GDC 50).

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 6.2.1.3, Rev 0: Mass and II.1 Compliance with GDC 50 and Departure The energy from metal-water reactions is 6.2.1 gy Release Analysis for 10 CFR 50, Appendix K < paragraph not included. The NuScale design supports ulated Loss-of-Coolant I.A - Sources of Heat during the LOCA an exemption from selected portions of dents (LOCAs) 10 CFR 50, Appendix K. S 6.2.1.3, Rev 0: Mass and II.2 Inspections, Tests, Analyses, and Conforms None. 6.2.1 gy Release Analysis for Acceptance Criteria (ITAAC) for ulated Loss-of-Coolant Design Certification Applications dents (LOCAs) S 6.2.1.3, Rev 0: Mass and II.3 Inspections, Tests, Analyses, and Not Applicable This acceptance criterion is applicable only Not Applicable gy Release Analysis for Acceptance Criteria (ITAAC) for to COL applicants. ulated Loss-of-Coolant Combined License (COL) Applications dents (LOCAs) S 6.2.1.4, Rev 0: Mass and II.1 Sources of Energy Conforms None. 6.2.1 gy Release Analysis for ulated Secondary System Ruptures S 6.2.1.4, Rev 0: Mass and II.2 Mass and Energy Release Rate Conforms None. 6.2.1 gy Release Analysis for ulated Secondary System Ruptures S 6.2.1.4, Rev 0: Mass and II.3 Single-Failure Analyses Conforms None. 6.2.1 gy Release Analysis for ulated Secondary System Ruptures 6.2.1.5, Rev 3: Minimum All Containment Pressure Model for Not Applicable This SRP section and its acceptance criteria Not Applicable Conformance with Regulatory Criteria ainment Pressure ECCS Performance Analysis; are applicable only to PWRs for which a ysis for Emergency Core Containment Response Analyses postulated LOCA results in core uncovery. ing System Performance Conservatism For the NuScale reactor design, a LOCA bility Studies does not result in core uncovery. S 6.2.2, Rev 0: II.1 GDC 5, Sharing of Structures, Conforms None. 6.2.2 ainment Heat Removal Systems, and Components 9.2.5 ems

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 6.2.2, Rev 0: II.2 GDC 38, Containment Heat Removal Departure The NuScale design supports an exemption 6.2.2 ainment Heat Removal from the power provisions of GDC 38. As ems described in Section 3.1.4, the design complies with a NuScale-specific principal design criterion in lieu of this GDC. S 6.2.2, Rev 0: II.3 GDC 39, Inspection of Containment Conforms None. 6.2.2 ainment Heat Removal Heat Removal System ems S 6.2.2, Rev 0: II.4 GDC 40, Testing of Containment Heat Departure The NuScale design does not conform to 3.1.4 ainment Heat Removal Removal System GDC 40 and the design supports an 6.2.2 ems exemption. S 6.2.2, Rev 0: II.5 10 CFR 50.46(b)(5), long-term cooling, Conforms None. 6.2.2 ainment Heat Removal including adequate water level (head) ems margin RRVs), in the presence of LOCA-generated and latent debris S 6.2.2, Rev 0: II.6 Compliance with 10 CFR 50.46(b)(5) Conforms None. 6.2.2 ainment Heat Removal as it relates to requirements for long-ems term cooling 6.2.3, Rev 3: Secondary All Various Not Applicable This SRP section and its acceptance criteria Not Applicable ainment Functional (II.1 through II.4) apply only to LWR designs gn that incorporate primary and secondary containment. The NuScale containment vessel design does not include a secondary containment. S 6.2.4, Rev 0: II.1 Instrument Line Isolation Conforms No instrumentation process lines penetrate 6.2.4 Conformance with Regulatory Criteria ainment Isolation containment. em S 6.2.4, Rev 0: II.2 Isolation of and Leak Detection in Conforms None. 6.2.4 ainment Isolation Lines in Engineered Safety Feature (or em Related) Systems S 6.2.4, Rev 0: II.3 Isolation of and Leak Detection in Conforms None. 6.2.4 ainment Isolation Lines in Systems Needed for Safe em Shutdown S 6.2.4, Rev 0: II.4 Containment Isolation Valve Conforms None. 6.2.4 ainment Isolation Requirements em

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 6.2.4, Rev 0: II.5 Containment Isolation Valve Conforms None. 6.2.4 ainment Isolation Requirements for Engineered Safety em Feature (or Related) Systems S 6.2.4, Rev 0: II.6 Use of Sealed-Closed Barriers in Place Conforms None. 6.2.4 ainment Isolation of Automatic Isolation Valves em S 6.2.4, Rev 0: II.7 Use of Relief Valves as Isolation Valves Not Applicable Relief valves are not used as CIVs. Not Applicable ainment Isolation em S 6.2.4, Rev 0: II.8 Classification of Essential or Conforms None. 6.2.4 ainment Isolation NonEssential Systems em S 6.2.4, Rev 0: II.9 Location of Isolation Valves Outside Conforms None. 6.2.4 ainment Isolation Containment em S 6.2.4, Rev 0: II.10 Loss of Power to Automatic Isolation Conforms None. 6.2.4 ainment Isolation Valves em S 6.2.4, Rev 0: II.11 Isolation Reliability Conforms None. 6.2.4 ainment Isolation em S 6.2.4, Rev 0: II.12 Parameter Diversity for Initiation of Conforms None. 6.2.4 ainment Isolation Containment Isolation em Conformance with Regulatory Criteria

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 6.2.4, Rev 0: II.13 Radiation Monitors for Initiation of Departure The containment evacuation system has the 6.2.4 ainment Isolation Containment Isolation on Open Paths potential for an open path from em to the Environs containment to the environs but is isolated upon a high containment vessel pressure signal, and a low-low pressurizer level signal. Any in-containment event resulting in core damage or degradation also results in containment isolation on low low pressurizer level and high containment pressure. Any event leading to core damage or degradation, results in containment isolation on low low pressurizer level. These features provide an alternative, reliable means to prevent radiological release from the CES to the environs, consistent with the intent of this Acceptance Criterion. The NuScale design supports an exemption from 50.34(f)(2)(xiv). S 6.2.4, Rev 0: II.14 Isolation Valve Closure Times Conforms None. 6.2.4 ainment Isolation em S 6.2.4, Rev 0: II.15 Use of Closed System Inside Conforms None. 6.2.4 ainment Isolation Containment em S 6.2.4, Rev 0: II.16 Specific Design Criteria for Conforms None. 6.2.4 Conformance with Regulatory Criteria ainment Isolation Containment Isolation Components em S 6.2.4, Rev 0: II.17 Provisions to Allow Control Room Conforms None. 6.2.4 ainment Isolation Operator Actions em S 6.2.4, Rev 0: II.18 Operability and Leakage Rate Testing Conforms None. 6.2.4 ainment Isolation em S 6.2.4, Rev 0: II.19 Reopening of Containment Isolation Conforms None. 6.2.4 ainment Isolation Valves em

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 6.2.4, Rev 0: II.20 Station Blackout Conforms None. 6.2.4 ainment Isolation em S 6.2.4, Rev 0: II.21 Source Term in Radiological Conforms None. 6.2.4 ainment Isolation Calculations em S 6.2.5, Rev 0: II.1 Analysis of Hydrogen and Oxygen Partially Conforms The NuScale design combustible gas 6.2.5 bustible Gas Control in Concentration Control and control system does not use combustible ainment Distribution in Containment gas control systems. Systems to control hydrogen concentrations within containment are not required because combustion has no impact on CNV integrity. The NuScale design supports an exemption to 10 CFR 50.44(c)(2) as described in DCA Part 7, section 2. The NuScale design includes hydrogen and oxygen monitoring equipment that is capable of continuously measuring the concentration of hydrogen and oxygen in the containment atmosphere following a significant beyond design-basis accident for accident management and emergency planning. S 6.2.5, Rev 0: II.2 Equipment Survivability and Conforms None. 6.2.5 bustible Gas Control in Containment Structural Integrity ainment Conformance with Regulatory Criteria S 6.2.5, Rev 0: II.3 Ensuring a Mixed Atmosphere Conforms None. 6.2.5 bustible Gas Control in ainment S 6.2.5, Rev 0: II.4 Design Requirements of GDC 41 Departure The NuScale design supports an exemption 6.2.5 bustible Gas Control in from the power provisions of GDC 41. As ainment described in Section 3.1.4, the design complies with a NuScale-specific principal design criterion in lieu of this GDC.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 6.2.5, Rev 0: II.5 Inspection and Test Requirements of Not Applicable For GDC 42 and 43, the NuScale design does Not Applicable bustible Gas Control in GDC 41, GDC 42, and GDC 43 not include a containment atmospheric ainment cleanup system. Containment integrity is assured without systems to control hydrogen and oxygen concentrations within containment. See acceptance criterion II.4 above for GDC 41 compliance. S 6.2.5, Rev 0: II.6 Containment Structural Integrity Conforms None. 6.2.5 bustible Gas Control in Analysis ainment S 6.2.6, Rev 0: All Various Departure The NuScale design supports an exemption 6.2.6 ainment Leakage from the containment leakage rate testing ng at design pressure requirements of GDC 52 and Type A test requirements of 10 CFR 50 Appendix J. 6.2.7, Rev 1: Fracture All Various Conforms None. 6.2.7 ention of Containment sure Boundary S 6.3, Rev 0: Emergency II.1 ECCS Acceptance Criteria of Conforms None. 6.3.1 Cooling System 10 CFR 50.46 6.3.3 S 6.3, Rev 0: Emergency II.2 Single-Failure Consideration Conforms None. 6.3.1 Cooling System S 6.3, Rev 0: Emergency II.3 Inservice Inspection and Operability Departure None. 6.3.2 Cooling System Testing Conformance with Regulatory Criteria S 6.3, Rev 0: Emergency II.4 Combined Reactivity Control System Departure The guidance in this acceptance criterion 6.3.1 Cooling System Capability and Actuation Provisions related to actuation signals is applicable to ECCS actuation. For the requirements of GDC 27, the NuScale ECCS does not perform a poison addition safety function nor does it provide a makeup function. The NuScale design supports an exemption to GDC 27. As described in Section 3.1.4, the design complies with a NuScale-specific principal design criterion in lieu of this GDC.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 6.3, Rev 0: Emergency II.5 Water Hammer Conforms None. 6.3.1 Cooling System S 6.3, Rev 0: Emergency II.6 Design of Non-Safety-Related Conforms None. 6.3.1 Cooling System Portions of ECCS S 6.3, Rev 0: Emergency II.7 ECCS Interfaces and Shared Systems Conforms None. 6.3.1 Cooling System S 6.3, Rev 0: Emergency II.8 Long Term Cooling Conforms None. 6.3.1 Cooling System S 6.3, Rev 0: Emergency II.9 ECCS Outage Times and Reports on Conforms None. 6.3.2 Cooling System Unavailability S 6.3, Rev 0: Emergency II.10 Programmatic Requirements Conforms None. 6.3.1 Cooling System 6.4, Rev 3: Control Room II.1 Control Room Emergency Zone Conforms None. 6.4 tability System 6.4, Rev 3: Control Room II.2 Ventilation System Criteria Conforms None. 6.4 tability System 6.4, Rev 3: Control Room II.3 Pressurization Systems Conforms None. 6.4 tability System 6.4, Rev 3: Control Room II.4 Emergency Standby Atmosphere Not Applicable This guidance is applicable only to reactor Not Applicable tability System Filtration System designs that rely on emergency filtration for control room habitability during a design basis accident. The NuScale control room habitability system neither relies on nor uses emergency filtration to protect Conformance with Regulatory Criteria operators during accident conditions. Rather, clean air is provided using compressed air tanks.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 6.4, Rev 3: Control Room II.5 Relative Location of Source and Not Applicable This guidance is applicable only to reactor Not Applicable tability System Control Room designs that rely on the control room emergency ventilation system for control room habitability during a design basis accident. The NuScale control room habitability system uses compressed air tanks as a clean air source during postulated accident events. This eliminates the potential for radioactive material or toxic gases to enter the control room via ventilation system inlets. 6.4, Rev 3: Control Room II.6.A Dose Guidelines for Current Not Applicable This guidance is applicable only to currently Not Applicable tability System Operating Reactors That Do Not operating reactors. Implement an Alternative Source Term 6.4, Rev 3: Control Room II.6.B Dose Guidelines for New Reactors Conforms The subtier RG 1.183 is partially applicable. 6.4.1 tability System and Licensees That Implement an Alternative Source Term 6.4, Rev 3: Control Room II.7 Toxic Gas Hazards Partially Conforms Programmatic requirements are the COL 6.4 tability System applicant responsibility. 6.5.1, Rev 4: ESF II First full paragraph on Page 6.5.1-6, Not Applicable The NuScale Power Plant design does not Not Applicable osphere Cleanup Design, Testing, and Maintenance of use engineered safety feature (ESF) filter ems ESF Atmosphere Cleanup System Air systems or ESF ventilation systems to Filtration and Adsorption Units mitigate the consequences of a design basis accident. In the NuScale Power Plant design Conformance with Regulatory Criteria there is a nonsafety-related Reactor Building heating ventilating and air conditioning system which includes filtering; however, it is not credited in the dose analysis. 6.5.2, Rev 4: Containment All Various Not Applicable This SRP section and its acceptance criteria Not Applicable y as a Fission Product (II.1 through II.3) are applicable only to large nup System LWRs with containment spray systems. The NuScale containment vessel design does not incorporate a spray system.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 6.5.3, Rev 3: Fission II.1 Primary Containment Partially Conforms A portion of this acceptance criterion and its 6.5.3 uct Control Systems and subtier guidance is applicable only to LWR ctures designs that include containment fission product clean-up systems. The NuScale containment vessel does not contain fission product clean up systems, nor does it include or require pressure suppression systems (e.g., suppression pools or active containment heat removal systems such as containment spray) that serve a fission product removal/dose mitigation function. Rather, fission product control is inherent in the passive design of the NuScale Power Module, wherein the compact containment vessel is submerged in the reactor pool. Therefore, the aspects of this guidance related to these systems are not applicable to the DCA. This guidance is applicable to the review of certain NuScale containment parameters and design features, such as design leakage rate and systems leakage prior to containment isolation. 6.5.3, Rev 3: Fission II.2 Secondary Containment Not Applicable This acceptance criterion is applicable only Not Applicable uct Control Systems and to LWRs that incorporate both a primary ctures and secondary containment. The NuScale Conformance with Regulatory Criteria containment vessel design does not include a secondary containment. 6.5.3, Rev 3: Fission II.4 Other Fission Product Control Not Applicable The only credited ESF fission product Not Applicable uct Control Systems and Systems control system in the NuScale Power Plant ctures design is the containment vessel in conjunction with the containment isolation valves and passive containment isolation barriers.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 6.5.4, Draft Rev 4: Ice All Various Not Applicable This SRP section and its acceptance criteria Not Applicable denser as a Fission are applicable only to applicants for plant uct Cleanup System designs that involve ice condenser containments. The NuScale reactor design does not use an ice condenser containment. 6.5.5, Rev 1: Pressure All Various Not Applicable This SRP section and its acceptance criteria Not Applicable ression Pool as a Fission are applicable only to large LWRs that credit uct Cleanup System a pressure suppression pool for fission product scrubbing and retention (i.e., BWRs). The NuScale reactor design does not credit or use a suppression pool. S 6.6, Rev 0: Inservice II.1 Components Subject to Inspection Conforms None. 6.6.1 ection and Testing of s 2 and 3 Components S 6.6, Rev 0: Inservice II.2 Accessibility Conforms None. 6.6.2 ection and Testing of s 2 and 3 Components S 6.6, Rev 0: Inservice II.3 Examination Categories and Methods Conforms None. 6.6.3 ection and Testing of s 2 and 3 Components S 6.6, Rev 0: Inservice II.4 Inspection Intervals Conforms None. 6.6.4 ection and Testing of s 2 and 3 Components S 6.6, Rev 0: Inservice II.5 Evaluation of Examination Results Conforms None. 6.6.5 ection and Testing of Conformance with Regulatory Criteria s 2 and 3 Components S 6.6, Rev 0: Inservice II.6 System Pressure Tests Conforms None. 6.6.7 ection and Testing of s 2 and 3 Components S 6.6, Rev. 0: Inservice II.7 Structural Supports Conforms None. 6.6.1 ection and Testing of 6.6.5 s 2 and 3 Components Table 6.6-1 S 6.6, Rev 0: Inservice II.8 Augmented ISI to Protect Against Conforms None. 6.6.8 ection and Testing of Postulated Piping Failures s 2 and 3 Components

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 6.6, Rev 0: Inservice II.9 Code Exemptions Conforms None. 6.6 ection and Testing of s 2 and 3 Components S 6.6, Rev 0: Inservice II.10 Relief Requests Conforms None. 6.6 ection and Testing of s 2 and 3 Components S 6.6, Rev 0: Inservice II.11 Code Cases Conforms None. 6.6 ection and Testing of s 2 and 3 Components S 6.6, Rev 0: Inservice II.12 Operational Programs Not Applicable The operational program and Not Applicable ection and Testing of implementation milestones governed by s 2 and 3 Components this acceptance criterion are the responsibility of the COL applicant. S 6.6, Rev 0: Inservice II.13 Risk Informed ISI Program Not Applicable None. Not Applicable ection and Testing of s 2 and 3 Components 6.7, Draft Rev 3: Main All Various Not Applicable This SRP section and its acceptance criteria Not Applicable m Isolation Valve are applicable only to BWRs. age Control System R) 6-1, (March 2007): pH for B.1 Minimum pH for Emergency Coolant Conforms This acceptance criterion is applicable but 6.1.1 rgency Coolant Water for Water certain language in BTP 6-1, which would surized Water Reactors be applied by Acceptance Criterion B.1, refers to SSC that are not in the NuScale design. Specifically, the NuScale design Conformance with Regulatory Criteria does not use a containment spray system or a sump. However, during ECCS operation, ECCS water does collect inside the NuScale containment vessel for recirculation back to the reactor core, and thus the intent of this acceptance criterion is applicable to the DCA.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 6-1, (March 2007): pH for B.2 Spray Water pH and Water Chemistry Partially Conforms The intent of a portion of this acceptance 6.2.2 rgency Coolant Water for Requirements for Fission Product criterion is applicable but the specific surized Water Reactors Removal language refers to SSC that are not in the NuScale design. Specifically the NuScale design does not use a containment spray system or a sump. However, during ECCS operation, ECCS water does collect inside the NuScale containment vessel for recirculation back to the reactor core, and thus the pH guideline contained in this acceptance criterion is applicable to the DCA. 6-1, (March 2007): pH for B.3 Hydrogen Generation from Conforms This acceptance criterion is applicable but 6.3.2 rgency Coolant Water for Aluminum Corrosion certain language in BTP 6-1, which would surized Water Reactors be applied by Acceptance Criterion B.3, refers to SSC that are not in the NuScale design. Specifically, the NuScale design does not use a containment spray system or a sump. However, during ECCS operation, ECCS water does collect inside the NuScale containment vessel for recirculation back to the reactor core, and thus the intent of this acceptance criterion is applicable to the DCA. 6-2, Rev 3: Minimum All (B.1 thru Various Not Applicable This guidance is applicable only to PWRs for Not Applicable Conformance with Regulatory Criteria ainment Pressure Model B.3) which a postulated LOCA results in core WR ECCS Performance uncovery. For the NuScale design, a LOCA does not result in core uncovery. 6-3, Rev 3: Determination All Various Not Applicable These acceptance criteria (B.1 through B.9) Not Applicable pass Leakage Paths in are applicable only to large LWRs that Containment Plants incorporate both a primary and secondary containment. The NuScale containment vessel design does not include a secondary containment.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 6-4, Rev 3: Containment All (B.1 thru Various Not Applicable This guidance pertains to containment Not Applicable ing During Normal Plant B.5) purge systems used to vent containment rations directly to the environs. While the NuScale containment vessel design includes an evacuation system, it serves a different purpose than a purge system, and includes features that provide suitable means to prevent radiological release to the environs (see DSRS 6.2.4, AC II.13). (The NuScale containment vessel evacuation system valve closure times are addressed under SRP Section 6.2.4.) 6-5, Rev 3: Currently the All Various Not Applicable This guidance is applicable only to LWR Not Applicable onsibility of Reactor ECCS designs that rely on safety injection ems Piping From the pumps and refueling (or borated) water T (or BWST) and storage tanks. The NuScale ECCS design ainment Sump(s) to the does not use pumps or refueling water ty Injection Pumps storage tanks (or equivalent). S 7.0, Rev 0: All Various Conforms This DSRS section provides a general 7.0 umentation and Controls description of the process for reviewing I&C oduction and Overview systems that is applicable to the DCA. view Process However, this guidance does not contain specific acceptance criteria. Specific acceptance criteria for SRP Chapter 7 are provided in the individual SRP Chapter 7 Conformance with Regulatory Criteria sections and are summarized in SRP Section 7.1, SRP Table 7-1, and SRP Appendix 7.1-A. S Appendix 7.0-A, Rev 0: - I&C - Hazard Analysis Conforms None. 7.1.8 Hazard Analysis S Appendix 7.0-B, Rev 0: - I&C - System Architecture Conforms None. 7.0.3 System Architecture 7.0.4 7.1 7.2 S Appendix 7.0-C, Rev 0: - I&C - Simplicity Conforms None. 7.1.6 Simplicity 7.1.7 7.1.8

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 7.1.1, Rev 0: All Specific SRP Acceptance Criteria Conforms None. 7.1 amental Design Applicable to I&C Systems Important 7.1.1 iples to Safety are Listed in Table 7.0-1. S 7.1.2, Rev 0: II.1 Ensure compliance to current version Conforms None. 7.1.2 pendence of RG 1.75 S 7.1.2, Rev 0: II.2 Ensure compliance to current version Conforms None. 7.1.2 pendence of RG 1.152 S 7.1.3, Rev 0: - Conformance with RG 1.53 Conforms None. 7.1.3 ndancy S 7.1.4, Rev 0: - Predictability and Repeatability Conforms None. 7.1.4 ictability and atability S 7.1.5, Rev 0: Diversity II.1 Methods for performing D3 analyses Conforms The D3 assessment of the NuScale I&C 7.1.5 Defense in Depth of reactor protection systems design is consistent with the guidelines in NUREG/CR-6303. S 7.1.5, Rev 0: Diversity II.2 SECY-93-087 Conforms Conformance to the applicable regulatory 7.1.5 Defense in Depth guidance from the staff requirement memorandum to SECY-93-087 is summarized in Section 7.1.5. S 7.1.5, Rev 0: Diversity II.3 GL 85-06 Conforms Conformance to 10 CFR 50.62 is 7.1.1 Defense in Depth summarized in Section 7.1.6. I&C systems 7.1.5 are designed to the quality assurance 7.1.6 program described in Section 17.5. S 7.1.5, Rev 0: Diversity II.4 Conformance to RG 1.53 Conforms None. 7.1.5 Conformance with Regulatory Criteria Defense in Depth S 7.1.5, Rev 0: Diversity II.5 Conformance to RG 1.62 Conforms See RG 1.62 in Table 1.9-2. 7.1.5 Defense in Depth S 7.1.5, Rev 0: Diversity II.6 Conformance to IEEE Std. 7-4.3.2 Conforms None. 7.1.1 Defense in Depth 7.1.2 7.1.5 S 7.2.1 Rev. 0: Quality II.1 Conformance to RG 1.28 Conforms See RG 1.28 in Table 1.9-2. 7.2.1 S 7.2.1 Rev. 0: Quality II.2 Conformance to RG 1.152 Conforms See RG 1.152 in Table 1.9-2. 7.2.1 S 7.2.1 Rev. 0: Quality II.3 Conformance to RG 1.168 Partially Conforms See RG 1.168 in Table 1.9-2. 7.2.1 S 7.2.1 Rev. 0: Quality II.4 Conformance to RG 1.169 Partially Conforms See RG 1.169 in Table 1.9-2. 7.2.1 S 7.2.1 Rev. 0: Quality II.5 Conformance to RG 1.170 Partially Conforms See RG 1.170 in Table 1.9-2. 7.2.1 S 7.2.1 Rev. 0: Quality II.6 Conformance to RG 1.171 Partially Conforms See RG 1.171 in Table 1.9-2. 7.2.1

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 7.2.1 Rev. 0: Quality II.7 Conformance to RG 1.172 Partially Conforms See RG 1.172 in Table 1.9-2. 7.2.1 S 7.2.1 Rev. 0: Quality II.8 Conformance to RG 1.173 Partially Conforms See RG 1.173 in Table 1.9-2. 7.2.1 S 7.2.2, Rev 0: Equipment II.1 Conformance to IEEE Std 7- 4.3.2 Conforms Digital I&C safety systems conform to the 7.2.2 ification guidance in Section 5.4 of IEEE Std 7- 4.3.2-2003, IEEE Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations, as endorsed (with identified exceptions and clarifications) by RG 1.152, Rev. 3. S 7.2.2, Rev 0: Equipment II.2 Conformance to RG 1.209 Conforms None. 7.2.2 ification S 7.2.2, Rev 0: Equipment II.3 Conformance to RG 1.151 Partially Conforms See RG 1.151 in Table 1.9-2. 7.2.2 ification S 7.2.2, Rev 0: Equipment II.4 Conformance to RG 1.180 Partially Conforms See RG 1.180 in Table 1.9-2. 7.2.2 ification S 7.2.2, Rev 0: Equipment II.5 Conformance to RG 1.204 Partially Conforms See RG 1.204 in Table 1.9-2. 7.2.2 ification S 7.2.3, Rev 0: Reliability, II.1 Conformance to IEEE Std 7-4.3.2 Conforms Digital I&C safety systems conform to the 7.2.3 grity, and Completion of reliability, integrity, and completion of ective Action protective action guidance in Sections 5.5, and 5.15 of IEEE Std 7-4.3.2-2003, as endorsed by RG 1.152 Rev. 3. S 7.2.4, Rev 0: Operating II.1 Conformance to RG 1.47 Conforms None. 7.2.4 Maintenance Bypasses Conformance with Regulatory Criteria S 7.2.5, Rev 0: Interlocks II.1 Conformance to IEEE Std 7-4.3.2 Conforms For computer-based interlocks, the 7.2.5 components and system conform to the guidance for digital computers in IEEE Std 7-4.3.2, as endorsed (with identified exceptions and clarifications) by RG 1.152 Rev. 3. S 7.2.6, Rev 0: Derivation All Various Conforms There are no specific DSRS acceptance 7.2.6 stem Inputs criteria in this section. S 7.2.7, Rev 0: Setpoints II.1 Conformance to RG 1.105 Partially Conforms See RG 1.105 in Table 1.9-2. 7.2.7

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 7.2.7, Rev 0: Setpoints II.2 NRC Regulatory Issue Summary (RIS) Conforms The setpoint methodology conforms to 7.2.7 2006-17 ISA-67.04.01-2006, Setpoints for Nuclear Safety-Related Instrumentation, which addresses the issues identified in RIS 2006-17. S 7.2.7, Rev 0: Setpoints II.3 Generic Letter (GL) 91-04 Conforms The guidance of GL 91-04 is applicable to 7.2.7 the setpoint methodology as described in TR-616-49121, NuScale Instrument Setpoint Methodology Technical Report (Ref. 7.2-27). S 7.2.8, Rev 0: Auxiliary All Various Conforms There are no specific DSRS acceptance 7.2.8 ures criteria in this section. S 7.2.9, Rev 0: Control of II.1 Conformance to IEEE Std 7-4.3.2 Conforms Digital I&C safety systems and components 7.2.9 ss, Identification, and conform to the identification guidance in ir Section 5.11 of IEEE Std 7-4.3.2-2003. S 7.2.9, Rev 0: Control of II.2 Conformance to RG 1.75 Conforms None. 7.2.9 ss, Identification, and ir S 7.2.10, Rev 0: All Varies Conforms There are no specific DSRS acceptance 7.2.10 action Between Sense criteria in this section. However, the Command Features and guidance provided is used to review the r Systems acceptability of the information associated with interaction between sense and command features and other systems. S 7.2.11, Rev 0: Multi-Unit II.1 Conformance to RG 1.53 Conforms None. 7.2.11 ons Conformance with Regulatory Criteria S 7.2.12, Rev 0: Automatic II.1 Conformance to RG 1.62 Conforms See RG 1.62 in Table 1.9-2. 7.2.12 Manual Controls S 7.2.13, Rev 0: Displays II.1 Conformance to RG 1.97 Partially Conforms See RG 1.97 in Table 1.9-2. 7.2.13 Monitoring S 7.2.13, Rev 0: Displays II.2 Conformance to RG 1.47 Conforms None. 7.2.4 Monitoring 7.2.13 7.2.15 S 7.2.13, Rev 0: Displays II.3 SECY-93-087 Conforms The main control room and remote 7.2.13 Monitoring shutdown station are designed to maintain alarm system reliability in accordance with item II.T of SECY-93-087.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 7.2.14, Rev 0: Human All Various Conforms There are no specific DSRS acceptance 7.2.14 ors Considerations criteria in this section. However, the guidance provided is used to review the acceptability of information associated with human factors considerations. S 7.2.15, Rev 0: Capability II.1 Conformance to IEEE Std 7-4.3.2 Conforms Digital I&C safety systems and components 7.2.15 est and Calibration conform to the guidance related to capability for test and calibration in Sections 5.7, 5.5.2, and 5.5.3 of IEEE Std 7-4.3.2-2003. S 7.2.15, Rev 0: Capability II.2 Conformance to RG 1.118 Partially Conforms See RG 1.118 in Table 1.9-2. 7.2.15 est and Calibration S 8.1, Rev 0: Electric II (No Number) Specific SRP Acceptance Criteria Partially Conforms DSRS Table 8-1 provides a matrix of the NRC 8.1.4 er - Introduction Contained in SRP Sections 8.2, 8.3.1, requirements, guidance, and Commission 8.2.2 8.3.2, and 8.4 (summarized in Table 8- policy documents, and industry codes and 8.3.1

1) standards that are applied as acceptance 8.3.2 criteria and guidance to the review of the 8.4 electrical systems described in Sections 8.2, 8.3.1, 8.3.2, and 8.4. Some of these documents are not relevant or are only partially relevant to the NuScale design.

S 8.2, Rev 0: Offsite Power II.1 Compliance with GDC 5 Not Applicable Conformance with GDC 5 is the Not Applicable em responsibility of the COL applicant as described in Section 8.2.2. S 8.2, Rev 0: Offsite Power II.2 Compliance with GDC 17 Departure The NuScale design supports an exemption 8.2.3 Conformance with Regulatory Criteria em from GDC 17 that includes the associated requirements for the offsite power system. S 8.2, Rev 0: Offsite Power II.3 Compliance with GDC 18 Departure The NuScale design supports an exemption 8.2.3 em from GDC 18 that includes the associated requirements for the offsite power system. S 8.2, Rev 0: Offsite Power II.4 Compliance with GDC 33 Departure The NuScale design supports an exemption 8.2.3 em from GDC 33. S 8.2, Rev 0: Offsite Power II.4 Compliance with GDCs 34, 35, 38, 41, Departure NuScale complies with a set of principal 8.2.3 em and 44 design criteria in lieu of these GDC.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 8.2, Rev 0: Offsite Power II.5 Compliance with 10 CFR 50.63 - Conforms The details regarding conformance with 8.2.3 em Passive Design 10 CFR 50.63 are described in Section 8.4, 8.4 Station Blackout. S 8.2, Rev 0: Offsite Power II.6 Compliance with 10 CFR 50.65(a)(4) Not Applicable Development of the maintenance rule Not Applicable em (10 CFR 50.65) program is the responsibility of the COL applicant referencing the certified design. S 8.3.1, Rev 0: AC Power II.1 Compliance with GDC 2 Conforms Onsite AC power systems conform to GDC 2 8.1.4 ems (Onsite) to the extent described in Section 8.3.1.2.1. 8.3.1 S 8.3.1, Rev 0: AC Power II.2 Compliance with GDC 4 Conforms Onsite AC power systems conform to GDC 4 8.1.4 ems (Onsite) to the extent described in Section 8.3.1.2.2. 8.3.1 S 8.3.1, Rev 0: AC Power II.3 Compliance with GDC 5 Partially Conforms Onsite AC power systems conform to GDC 5 8.1.4 ems (Onsite) to the extent described in Section 8.3.1. 8.3.1 S 8.3.1, Rev 0: AC Power II.4 Compliance with GDC 17 Departure The NuScale design supports an exemption 8.3.1 ems (Onsite) from GDC 17 that includes the associated requirements for the onsite AC power system. S 8.3.1, Rev 0: AC Power II.5 Compliance with GDC 18 Departure The NuScale design supports an exemption 8.1.4 ems (Onsite) from GDC 18 that includes the associated 8.3.1 requirements for the onsite AC power system. S 8.3.1, Rev 0: AC Power II (No Number) Compliance with GDC 33 Departure The NuScale design supports an exemption 8.1.4 ems (Onsite) from GDC 33. 8.3.1 S 8.3.1, Rev 0: AC Power II (No Number) Compliance with GDCs 34, 35, 38, 41, Departure NuScale complies with a set of principal 8.1.4 Conformance with Regulatory Criteria ems (Onsite) and 44 design criteria in lieu of these GDC. 8.3.1 S 8.3.1, Rev 0: AC Power II.6 Compliance with GDC 50 Conforms The electrical design requirements 8.1 ems (Onsite) associated with GDC 50 for electrical 8.3 penetration assemblies (EPAs) are included in Section 8.3. S 8.3.1, Rev 0: AC Power II.7 Compliance with 10 CFR 50.65(a)(4) Not Applicable Development of the maintenance rule ems (Onsite) (10 CFR 50.65) program including the identification of SSC that require assessment per 10 CFR 50.65(a)(4) is the responsibility of the COL applicant referencing the certified design.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 8.3.1, Rev 0: AC Power II.8 Compliance with 10 CFR 50.55a(h) Not Applicable No onsite electrical AC power system Not Applicable ems (Onsite) equipment is required to conform to 10 CFR 50.55a(h) and IEEE Std. 603-1991. S 8.3.1, Rev 0: AC II.9 Compliance with 10 CFR 52.47(b)(1) Conforms None. 8.1 er Systems (Onsite) 8.3 S 8.3.2, Rev 0: DC Power II (No Number) Compliance with GDC 2 Conforms None. 8.3.2 ems (Onsite) S 8.3.2, Rev 0: DC Power II (No Number) Compliance with GDC 4 Conforms None. 8.3.2 ems (Onsite) S 8.3.2, Rev 0: DC Power II (No Number) Compliance with GDC 5 Conforms None. 8.3.2 ems (Onsite) S 8.3.2, Rev 0: DC Power II (No Number) Compliance with GDC 17 Departure The NuScale design supports an exemption 8.3.2 ems (Onsite) from GDC 17 that includes the associated requirements for the onsite DC power systems. S 8.3.2, Rev 0: DC Power II (No Number) Compliance with GDC 18 Departure The NuScale design supports an exemption 8.3.2 ems (Onsite) from GDC 18 that includes the associated requirements for the onsite DC power systems. S 8.3.2, Rev 0: DC Power II (No Number) Compliance with GDC 33 Departure The NuScale design supports an exemption 8.3.2 ems (Onsite) from GDC 33. S 8.3.2, Rev 0: DC Power II (No Number) Compliance with GDC 34, 35, 38, 41, Departure Nuscale complies with a set of principal 8.3.2 ems (Onsite) and 44 design criteria in lieu of these GDC. S 8.3.2, Rev 0: DC Power II (No Number) Compliance with GDC 50 Conforms The electrical design requirements 8.1 Conformance with Regulatory Criteria ems (Onsite) associated with GDC 50 for electrical 8.3 penetration assemblies (EPAs) are included in Section 8.3. S 8.3.2, Rev 0: DC Power II.1 Conformance with RG 1.32 Partially Conforms As it applies to certain aspects of the EDSS 8.3.2 ems (Onsite) design. S 8.3.2, Rev 0: DC Power II.2 Conformance with RG 1.75 Conforms As it applies to certain aspects of the EDSS 8.3.2 ems (Onsite) design. S 8.3.2, Rev 0: DC Power II.3 Conformance with RG 1.81 Partially Conforms As it applies to certain aspects of the EDSS 8.3.2 ems (Onsite) design. S 8.3.2, Rev 0: DC Power II.4 Conformance with RG 1.118 Partially Conforms As it applies to certain aspects of the EDSS 8.3.2 ems (Onsite) design.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 8.3.2, Rev 0: DC Power II.5 Conformance with RG 1.153 Partially Conforms As it applies to certain aspects of the EDSS 8.3.2 ems (Onsite) design. S 8.3.2, Rev 0: DC Power II.6 Conformance with RG 1.153 Partially Conforms As it applies to certain aspects of the EDSS 8.3.2 ems (Onsite) design. S 8.3.2, Rev 0: DC Power II.7 Conformance with RG 1.63 Partially Conforms See RG 1.63 in Table 1.9-2. 8.1 ems (Onsite) 8.3.2 S 8.3.2, Rev 0: DC Power II.8 Conformance with RG 1.160 Not Applicable Development of the maintenance rule (10 8.3.2 ems (Onsite) CFR 50.65) program including the identification of SSC that require assessment per 10 CFR 50.65(a)(4) is the responsibility of the COL applicant referencing the certified design. S 8.4, Rev 0: Station II.1 Compliance with 10 CFR 50.63 and Partially Conforms None. 8.4 kout the guidelines of RG 1.155 S 8.4, Rev 0: Station II.2 Use of Alternate AC Power Sources Partially Conforms As described in Section 8.4, all safety- 8.4 kout and RTNSS for Plants of Passive related functions can be performed without 19.3 Design reliance on AC power for 72 hours after an SBO event. As described in Section 19.3, a RTNSS process has been implemented. Consequently, the Alternate AC Power Source is not applicable to the NuScale design. S 8.4, Rev 0: Station II.3 Independence of SBO-related power Partially Conforms Although DC power supplies are not 8.3.2 kout sources required to meet the SBO mitigation 8.4.3 requirements of 10 CFR 50.63, the Conformance with Regulatory Criteria independence of SBO related power supplies (EDSS) is described in Section 8.3. Appendix 8-A, Rev1: All Various Not Applicable This SRP appendix governs staff visits to Not Applicable eral Agenda, Station Site plant sites as part of licensing reviews s during the operating or COL stage. BTP 8-1, Rev 3: All (B.1 thru Various Not Applicable The NuScale design does not use safety Not Applicable irements on Motor- B.4) injection tanks (or equivalent) in response rated Valves in the ECCS to a design basis accident. Design and mulator Lines operation of the NuScale ECCS also do not involve motor-operated valves.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status BTP 8-2, Rev 3: Use of B. Use of Onsite Emergency Power Conforms The backup diesel generators are used only 8.1.1 el-Generator Sets for Diesel-Generator Sets for Purposes for supplying standby power to designated 8.3.1 ing Other Than Supplying Standby Power loads when needed, and are not is Prohibited interconnected with other AC power sources except for short periods for the purpose of load testing. BTP 8-3, Rev 3: Stability of B.1 Grid Reliability Not Applicable The analysis of grid stability is the Not Applicable te Power Systems responsibility of the COL applicant that references the NuScale design certification. BTP 8-3, Rev 3: Stability of B.2 Grid Capacity Not Applicable The analysis of grid stability is the Not Applicable te Power Systems responsibility of the COL applicant that references the NuScale design certification. BTP 8-4, Rev 3: All (B.1 Various Not Applicable BTP 8-4 establishes the acceptability of Not Applicable ication of the Single through B.5) disconnecting power to electrical re Criterion to Manually components of a fluid system as one means rolled Electrically of designing against a single failure that rated Valves might cause an undesirable component action. Removal of electric power from safety-related valves is not used in the design as a means of satisfying the single failure criterion. BTP 8-5, Rev 3: All (B.1 thru Design Criteria Reflecting Importance Not Applicable This BTP does not apply to NuScale electric Not Applicable plemental Guidance for B.6) of Providing Accurate Information to power systems as these systems are not ss and Inoperable Status the Operator and Reducing the engineered safety features and are not ation for Engineered Possibility of Adversely Affecting relied on to support engineered safety Conformance with Regulatory Criteria ty Features Systems Monitored Safety Systems features. BTP 8-6, Rev 3: Adequacy All Criteria for evaluating voltage Not Applicable For the NuScale design, the offsite power Not Applicable ation Electric Distribution protection for the offsite power system does not supply power to Class 1E em Voltages system to assure proper operation loads and does not support safety-related and sequencing of Class 1E loads functions. BTP 8-7, Rev 3: Criteria for All Design Criteria Reflecting Importance Not Applicable The NuScale plant does not require or Not Applicable ms and Indications of Providing Accurate Information to include safety-related emergency diesel ciated with Diesel- the Operator and Reducing the generators. erator Unit Bypassed and Possibility of Adversely Affecting erable Status Monitored Safety Systems

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status BTP 8-8, (Feb 2012): All Various Not Applicable With the nonreliance on AC power for Not Applicable te (Emergency Diesel safety-related functions, the operating erators) and Offsite restrictions (i.e., Technical Specifications er Sources Allowed Allowed Outage Times) for inoperable AC ge Time Extensions power sources specified in this guidance are not appropriate to apply. BTP 8-9, Rev. 0: Open B.1 Electrical system design to address Partially Conforms None. 8.2.3 e Conditions in Electric open phase condition er System BTP 8-9, Rev. 0: Open B.2 Criteria for evaluating open phase Not Applicable Not applicable to passive plant designs. Not Applicable e Conditions in Electric conditions for active plant designs er System BTP 8-9, Rev. 0: Open B.3 Criteria for evaluating open phase Partially Conforms None. 8.2.3 e Conditions in Electric conditions for passive plant designs er System 9.1.1, Rev 3: Criticality II.1 Specific Criteria to Meet GDC 62 Conforms None. 9.1.1.3 ty of Fresh and Spent 9.1.1.1 Storage and Handling S 9.1.2, Rev 0: New and II.1 Specific Criteria to Meet GDC 2 Conforms None. 9.1.2.1 t Fuel Storage 9.1.2.3 S 9.1.2, Rev 0: New and II.2 Specific Criteria to Meet GDC 4 Conforms None. 9.1.2.1 t Fuel Storage 9.1.2.3 S 9.1.2, Rev 0: New and II.3 Specific Criteria to Meet GDC 5 Conforms None. 9.1.2.1 t Fuel Storage 9.1.2.3 Conformance with Regulatory Criteria S 9.1.2, Rev 0: New and II.4 Specific Criteria to Meet GDC 61 Conforms An ESF ventilation system is not required 9.1.2.1 t Fuel Storage (see RG 1.52). 9.1.2.3 S 9.1.2, Rev 0: New and II.5 Specific Criteria to Meet GDC 63 Conforms None. 9.1.2.1 t Fuel Storage 9.1.2.3 9.1.2.5 S 9.1.2, Rev 0: New and II.6 Specific Criteria to Meet Conforms None. 9.1.2.1 t Fuel Storage 10 CFR 20.1101(b) 9.1.2.2 9.1.2.3 S 9.1.2, Rev 0: New and II.7 Criticality Monitors and Subcriticality Conforms None. 9.1.1 t Fuel Storage Margin

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 9.1.3, Rev 0: Spent Fuel II.1 Specific Criteria to Meet GDC 2 Partially Conforms The design conforms except that: (1) The 9.1.3.1 Cooling and Cleanup normal makeup water supply system and its 9.1.3.2 em source are not seismic Category I and the 9.1.3.3 system is not designed to Quality Group C per RG 1.26. The ultimate heat sink (UHS) system is a seismic Category I supply system and source for spent fuel cooling and shielding for accident conditions. A UHS makeup supply line is designed to Quality Group C and seismic Category I requirements. (2) An ESF ventilation system is not required (see RG 1.52). S 9.1.3, Rev 0: Spent Fuel II.2 Specific Criteria to Meet GDC 4 Partially Conforms This design conforms except that: (1) The 9.1.3.1 Cooling and Cleanup normal makeup water supply system and its 9.1.3.3 em source are not designed to accommodate the effects of postulated accidents. The UHS system is the supply system and source for spent fuel cooling and shielding that are designed to accommodate the effects of postulated accidents. A UHS makeup supply line is designed to meet GDC 4. (2) An ESF ventilation system is not required (see RG 1.52). S 9.1.3, Rev 0: Spent Fuel II.3 Specific Criteria to Meet GDC 5 Conforms None. 9.1.3.1 Cooling and Cleanup 9.1.3.3 Conformance with Regulatory Criteria em S 9.1.3, Rev 0: Spent Fuel II.4 Specific Criteria to Meet GDC 61 Conforms None. 9.1.3.1 Cooling and Cleanup 9.1.3.2 em S 9.1.3, Rev 0: Spent Fuel II.5 Specific Criteria to Meet GDC 63 Conforms None. 9.1.3.1 Cooling and Cleanup 9.1.3.5 em S 9.1.3, Rev 0: Spent Fuel II.6 Specific Criteria to Meet Conforms None. 9.1.3.1 Cooling and Cleanup 10 CFR 20.1101(b) 9.1.3.3 em

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 9.1.3, Rev 0: Spent Fuel II.7 ITAAC for Design Certification Conforms None. 14.3 Cooling and Cleanup Applications em S 9.1.3, Rev 0: Spent Fuel II.8 ITAAC for Combined License Not Applicable This acceptance criterion is applicable only Not Applicable Cooling and Cleanup Applications to COL applicants. em 9.1.4, Rev 4: Light Load II.1 Specific Criteria to Meet GDC 2 Conforms None. 9.1.4 dling System and eling Cavity Design 9.1.4, Rev 4: Light Load II.2 Specific Criteria to Meet GDC 5 Conforms None. 9.1.4 dling System and eling Cavity Design 9.1.4, Rev 4: Light Load II.3 Specific Criteria to Meet GDC 61 Conforms None. 9.1.4 dling System and eling Cavity Design 9.1.4, Rev 4: Light Load II.4 Specific Criteria to Meet GDC 62 Conforms None. 9.1.4 dling System and eling Cavity Design 9.1.5, Rev 1: Overhead II.1 Specific Criteria to Meet GDC 1 Conforms None. 9.1.5 y Load Handling ems 9.1.5, Rev 1: Overhead II.2 Specific Criteria to Meet GDC 2 Conforms None. 9.1.5 y Load Handling ems Conformance with Regulatory Criteria 9.1.5, Rev 1: Overhead II.3 Specific Criteria to Meet GDC 4 Conforms None. 9.1.5 y Load Handling ems 9.1.5, Rev 1: Overhead II.4 Specific Criteria to Meet GDC 5 Conforms None. 9.1.5 y Load Handling ems

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.2.1, Rev 5: Station II.1 Protection Against Natural Conforms The NuScale site cooling water system 9.2.7 ice Water System Phenomena (GDC 2) (SCWS) does not provide essential cooling (Used for Site to safety-related SSC and is not safety- Cooling Water related or important-to-safety. The System (SCWS)) applicability of GDC 2 to the SCWS reviewed under this acceptance criterion is limited to aspects ensuring that a failure of the nonsafety-related SCWS does not result in an adverse effect on a Seismic Category I SSC. For the NuScale design, this is provided by the design and construction of the nonsafety related SCWS to meet the provisions of RG 1.29, Staff Regulatory Guidance C.1.i. 9.2.1, Rev 5: Station II.2 Environmental and Dynamic Effects Partially Conforms The NuScale site cooling water system does 9.2.7 ice Water System (GDC 4) not provide essential cooling to safety- (Used for Site related SSC and is not considered safety- Cooling Water related or risk-significant. The applicability System (SCWS)) of GDC 4 to the NuScale cooling water system reviewed under this acceptance criterion is limited to aspects ensuring that a failure of the nonsafety-related SSC does not result in an adverse effect on a safety-related SSC. 9.2.1, Rev 5: Station II.3 Sharing of Structures, Systems, and Conforms The NuScale site cooling water system does 9.2.7 Conformance with Regulatory Criteria ice Water System Components (GDC 5) not provide essential cooling to safety- (Used for Site related SSC and are not safety-related or Cooling Water risk-significant. The design and layout of System (SCWS)) these systems satisfy GDC 5. Specifically, sharing of the site cooling water system between units has no reasonable likelihood of adversely affecting essential SSC and associated safety functions. 9.2.1, Rev 5: Station II.4 Cooling Water System (GDC 44) Not Applicable The site cooling water system does not Not Applicable ice Water System serve a safety-related cooling or accident mitigation function.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.2.1, Rev 5: Station II.5 Cooling Water System Inspection Not Applicable The site cooling water system does not Not Applicable ice Water System (GDC 45) serve a safety-related cooling or accident mitigation function. 9.2.1, Rev 5: Station II.6 Cooling Water System Testing (GDC Not Applicable The site cooling water system does not Not Applicable ice Water System 46) serve a safety-related cooling or accident mitigation function. 9.2.2, Rev 4: Reactor II.1 Protection Against Natural Partially Conforms The system function contemplated by this 9.2.2 liary Cooling Water Phenomena SRP criterion is applicable to the NuScale em reactor component cooling water (RCCW) system. This criterion is based on RG 1.29 position C.1 for safety-related portions, and position C.2 for nonsafety-related portions. Position C.1 is not applicable since the RCCW is not safety related. The NuScale RCCW complies with position C.2 in that the SSCs whose structural failure could affect the operability of safety-related SSCs are designed as Seismic Category II. 9.2.2, Rev 4: Reactor II.2 Environmental and Dynamic Effects Partially Conforms Additional information pertaining to impact 9.2.2 liary Cooling Water of environmental and dynamic effects is em provided in Sections 3.5 and 3.6. 9.2.2, Rev 4: Reactor II.3 Sharing of Structures, Systems, and Not Applicable See comment above for Acceptance Not Applicable liary Cooling Water Components Criterion II.1. em 9.2.2, Rev 4: Reactor II.4 Cooling Water System Not Applicable See comment above for Acceptance Not Applicable Conformance with Regulatory Criteria liary Cooling Water Criterion II.1. em 9.2.2, Rev 4: Reactor II.5 Cooling Water System Inspection Not Applicable See comment above for Acceptance Not Applicable liary Cooling Water Criterion II.1. em 9.2.2, Rev 4: Reactor II.6 Cooling Water System Testing Not Applicable See comment above for Acceptance Not Applicable liary Cooling Water Criterion II.1. em

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.2.4, Rev 3: Potable and II.1 Control of Releases of Radioactive Partially Conforms The NuScale potable and sanitary water 9.2.4 tary Water Systems Materials to the PWSW systems do not interface with systems potentially containing radioactivity. The NuScale potable and sanitary water systems are designed such that failure will not result in flooding or other adverse impacts on essential SSC. 9.2.5, Rev 3: Ultimate II.1 Protection Against Natural Partially Conforms Since RG 1.27 is not applicable to the 9.2.5 Sink Phenomena NuScale design, compliance with GDC 2 is demonstrated by adherence to RG 1.13, Regulatory Positions C.1 and C.2. The NuScale UHS provides both spent fuel cooling and containment heat removal, and is protected from natural phenomena and site-related events by the Seismic Category I RXB structure and with a Seismic Category I emergency makeup line. 9.2.5, Rev 3: Ultimate II.2 Sharing of Structures, Systems, and Conforms None. 9.2.5 Sink Components 9.2.5, Rev 3: Ultimate II.3 Cooling Water System Partially Conforms This acceptance criterion is applicable 9.2.5 Sink except for aspects related to the use of fiberglass piping (see RG 1.72). The NuScale design does not use fiberglass piping. 9.2.5, Rev 3: Ultimate II.4 Cooling Water System Inspection Conforms None. 9.2.5 Sink Conformance with Regulatory Criteria 9.2.5, Rev 3: Ultimate II.5 Cooling Water System Testing Conforms None. 9.2.5 Sink

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.2.6, Rev 3: Condensate II.1 Protection Against Natural Conforms The NuScale designs condensate storage 9.2.6 age Facilities Phenomena system is neither safety-related nor risk- 10.4.7 significant. The condensate storage systems and components are located outside the Seismic Category I Reactor Building. The effects of discharging water from a condensate storage facility failure have no reasonable potential to adversely impact the operation of safety-related systems or safe operation of the plant. Consistent with Staff Regulatory Guidance C.1.i of RG 1.29, no portion of the NuScale condensate storage system requires design and construction to withstand the safe-shutdown earthquake to prevent a failure that could adversely affect a Seismic Category I SSC. 9.2.6, Rev 3: Condensate II.2 Environmental and Dynamic Effects Conforms None. 9.2.6 age Facilities Design Basis 10.4.7 9.2.6, Rev 3: Condensate II.3 Sharing of Structures, Systems, and Conforms Sharing of the condensate storage facilities 9.2.6 age Facilities Components does not impair the ability of safety-related 10.4.7 or risk-significant SSC to perform their safety functions. 9.2.6, Rev 3: Condensate II.4 Control of Radioactive Releases to the Conforms None. 9.2.6 age Facilities Environment 10.4.7 Conformance with Regulatory Criteria 9.2.6, Rev 3: Condensate II.5 10 CFR 20.1406 Compliance Conforms None. 9.2.6 age Facilities 10.4.7 9.2.7, Rev 0: Chilled Water II.1 Quality Standards and Records Not Applicable The NuScale CHWS does not perform safety Not Applicable em or containment isolation functions. 9.2.7, Rev 0: Chilled Water II.2 Protection Against Natural Conforms This criterion is based on RG 1.29. The CHWS 9.2.8 em Phenomena is not classified as Seismic Category I. The CHWS complies with Staff Regulatory Guidance C.1.i in that the SSC whose failure could adversely affect Seismic Category I SSC are designed as Seismic Category II. 9.2.7, Rev 0: Chilled Water II.3 Environmental and Dynamic Effects Conforms None. 9.2.8 em

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.2.7, Rev 0: Chilled Water II.4 Sharing of Structures, Systems, and Not Applicable The NuScale CHWS is a nonsafety-system Not Applicable em Components and does not perform safety functions. 9.2.7, Rev 0: Chilled Water II.5 Cooling Water System Not Applicable The NuScale CHWS is a nonsafety-system Not Applicable em and does not perform safety functions. 9.2.7, Rev 0: Chilled Water II.6 Cooling Water System Inspection Not Applicable The NuScale CHWS is a nonsafety-system Not Applicable em and does not perform safety functions. 9.2.7, Rev 0: Chilled Water II.7 Cooling Water System Testing Not Applicable The NuScale CHWS is a nonsafety-system Not Applicable em and does not perform safety functions. 9.2.7, Rev 0: Chilled Water II.8 Minimization of Contamination Conforms The CHWS is at a higher pressure than the 9.2.8 em LRWS and GRWS where the systems interface, precluding introduction of radioactive contaminants into the CHWS. 9.3.1, Rev 2: Compressed II.1 Specific Criteria to Meet GDC 1 Not Applicable NuScale compressed air systems are non- Not Applicable ystem safety, non-risk-significant systems. 9.3.1, Rev 2: Compressed II.2 Specific Criteria to Meet GDC 2 Not Applicable NuScale compressed air systems are non- Not Applicable ystem safety, non-risk-significant systems. 9.3.1, Rev 2: Compressed II.3 Specific Criteria to Meet GDC 5 Conforms None. 9.3.1 ystem 9.3.1, Rev 2: Compressed II.4 Specific Criteria to Meet 10 CFR 50.63 Partially Conforms The intent of this acceptance criterion and 9.3.1 ystem its subtier guidance - to maintain the ability 8.4 to withstand and recover from a SBO lasting a specified minimum duration - are applicable. However, language that refers to reactor plant designs such as large LWRs Conformance with Regulatory Criteria is not relevant to the NuScale plant design. The NuScale plant design meets the intent of this guidance with its passive design and reduced reliance on AC power to cope with design-basis events. Specifically, compressed air is not required to achieve core cooling in the event of a station blackout in the NuScale design.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.3.2, Rev 3: Process and II.1 Sampling Capability Partially Conforms This acceptance criterion is applicable 9.3.2

-Accident Sampling                                                                              except for aspects that are BWR-specific, or ems                                                                                             not part of the NuScale design (e.g.,

refueling water storage tank, pressurizer relief tank, and containment sump). 9.3.2, Rev 3: Process and II.2 Technical Specifications Not Applicable This was addressed in NRC-approved TSTF Not Applicable

-Accident Sampling                                                                              366-A and is no longer applicable.

ems 9.3.2, Rev 3: Process and II.3 Process Sampling System Functional Conforms None. 9.3.2

-Accident Sampling                      Design ems 9.3.2, Rev 3: Process and   II.4        Seismic Design and Quality Group         Conforms        None.                                               9.3.2
-Accident Sampling                      Classification ems 9.3.3, Rev 3: Equipment     II.1        Protection Against Natural               Conforms        None.                                               9.3.3 Floor Drainage System                   Phenomena 9.3.3, Rev 3: Equipment     II.2        Environmental and Dynamic Effects        Conforms        None.                                               9.3.3 Floor Drainage System 9.3.3, Rev 3: Equipment     II.3        Control of Releases of Radioactive       Conforms       No portions of the NuScale drain system              9.3.3 Floor Drainage System                   Material to the Environment                             penetrate the containment barrier.

S 9.3.4, Rev 0: Chemical II.1 CVCS Functional Performance during Partially Conforms The only CVCS safety-related function 9.3.4 Volume Control System Adverse Environmental Phenomena; precludes inadvertent boron dilution of the R) (Including Boron Pumping Capacity; and defense-in- reactor coolant system. very System) depth RCS makeup Conformance with Regulatory Criteria S 9.3.4, Rev 0: Chemical II.2 Single Failure Criteria and GDC 5 Conforms The single-failure criteria apply only to the 9.3.4 Volume Control System two safety-related demineralized water R) (Including Boron isolation valves provided to preclude an very System) inadvertent boron dilution of the reactor coolant system. S 9.3.4, Rev 0: Chemical II.3 Minimization of contamination Conforms None. 9.3.4 Volume Control System R) (Including Boron very System)

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 9.3.4, Rev 0: Chemical II.4 Components of the RCPB, quality Conforms The CVCS is located outside the RCPB. 9.3.4 Volume Control System classification and seismic design R) (Including Boron classification very System) S 9.3.4, Rev 0: Chemical II.5 Chemical and Volume Control System Conforms None. 9.3.4 Volume Control System Design and Arrangement R) (Including Boron very System) S 9.3.4, Rev 0: Chemical II.6 Detection of Reactor Coolant Leakage Conforms None. 9.3.4 Volume Control System Outside Containment R) (Including Boron very System) S 9.3.4, Rev 0: Chemical II.7 Prevention of CVCS Holdup Tank Wall Partially Conforms A portion of this acceptance criterion is 9.3.4 Volume Control System Buckling/Failure; CVCS Venting and applicable but the specific language refers R) (Including Boron Draining to CVCS designs that are not relevant to the very System) NuScale design. The NuScale CVCS design does not have holdup tanks that are subject to the vacuum conditions in subtier Bulletin 80-05. The last sentence of this acceptance criterion is applicable to the NuScale CVCS design, which will include appropriate venting and draining capability. S 9.3.4, Rev 0: Chemical II.8 ITAAC Conforms None. 9.3.4 Volume Control System 14.3 R) (Including Boron Conformance with Regulatory Criteria very System) 9.3.5, Rev 3: Standby All Various Not Applicable This SRP section and its acceptance criteria Not Applicable id Control System (BWR) are applicable only to BWRs. S 9.3.6, Rev 0: II.1 GDC 2 Conforms None. 9.3.6 ainment Evacuation and ding Systems S 9.3.6, Rev 0: II.2 GDC 60 Conforms None. 9.3.6 ainment Evacuation and ding Systems

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 9.3.6, Rev 0: II.3 TMI 10 CFR 50.34(f) Conforms None. 9.3.6 ainment Evacuation and ding Systems 9.4.1, Rev 3: Control II.1 Protection Against Natural Conforms None. 9.4.1 m Area Ventilation Phenomena em 9.4.1, Rev 3: Control II.2 Environmental and Dynamic Effects Conforms None. 9.4.1 m Area Ventilation em 9.4.1, Rev 3: Control II.3 Sharing of Structures, Systems, and Conforms Operation of the CRVS is part of normal 9.4.1 m Area Ventilation Components plant operations. Up to 12 Modules em modules share the same control room. 9.4.1, Rev 3: Control II.4 Control Room Conforms None. 9.4.1 m Area Ventilation em 9.4.1, Rev 3: Control II.5 Control of Releases of Radioactive Conforms This acceptance criterion is applicable 9.4.1 m Area Ventilation Material to the Environment except for aspects related to ESF em atmosphere cleanup systems. The NuScale control room habitability system neither relies on nor uses emergency filtration to protect operators during accident conditions. Rather, clean air is provided using compressed air tanks. Conformance with Regulatory Criteria

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.4.1, Rev 3: Control II.6 Loss of All Alternating Current Power Conforms The intent of this acceptance criterion and 9.4.1 m Area Ventilation its subtier guidance - to maintain the ability em to withstand and recover from a station blackout (SBO) lasting a specified minimum duration - is applicable. However, much of the specific language refers to reactor plant designs such as large LWRs, and is not relevant to the NuScale plant design. The NuScale plant design meets the intent of this guidance with its passive design and reduced reliance on AC power to cope with design basis events. Consistent with Commission policy, this coping capability eliminates safety benefit a typical large LWR gains by having an alternate AC power source (e.g., gas turbine generator) for station blackout. Moreover, and specific to this SRP Section 9.4.1 acceptance criterion, the control room habitability system (Section 6.4) relies on compressed air tanks to pressurize the control room envelope in the event of an SBO. The design of the main control room and the surrounding walls, ceiling, and structure act as a passive heat sink to maintain the Conformance with Regulatory Criteria environment within acceptable conditions in the event of an SBO. 9.4.2, Rev 3: Spent Fuel II.1 Compliance with GDC 2 Conforms None. 9.4.2 Area Ventilation System 9.4.2, Rev 3: Spent Fuel II.2 Compliance with GDC 5 Conforms None. 9.4.2 Area Ventilation System 9.4.2, Rev 3: Spent Fuel II.3 Compliance with GDC 60 Conforms This acceptance criterion is applicable 9.4.2 Area Ventilation System except for aspects related to ESF atmosphere cleanup systems (see RG 1.52).

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.4.2, Rev 3: Spent Fuel II.4 Compliance with GDC 61 Conforms This acceptance criterion is applicable 9.4.2 Area Ventilation System except for aspects related to ESF atmosphere cleanup systems, as described in the comment above for RG 1.52 subtier to Acceptance Criterion II.3. 9.4.3, Rev 3: Auxiliary and II.1 Compliance with GDC 2 Conforms None. 9.4.3 waste Area Ventilation em 9.4.3, Rev 3: Auxiliary and II.2 Compliance with GDC 5 Conforms None. 9.4.3 waste Area Ventilation em 9.4.3, Rev 3: Auxiliary and II.3 Compliance with GDC 60 Not Applicable The RWBV system does not filter exhaust. Not Applicable waste Area Ventilation Exhaust is filtered by the RBV system. em 9.4.4, Rev 3: Turbine Area All (II.1 thru Compliance with GDC 2, GDC 5, and Not Applicable This SRP section and its acceptance criteria Not Applicable ilation System II.3) GDC 60 (II.1 through II.3) are applicable only to LWR designs that rely on the turbine area ventilation system, or portions thereof, to fulfill safety-related or risk-significant functions. The NuScale Turbine Building HVAC system (TBVS) is not relied on to control airborne radioactivity concentrations in the Turbine Building and gaseous effluents during normal operations (including anticipated operational Conformance with Regulatory Criteria occurrences) and after accidents that result in a radioactive material release. Furthermore, there are no requirements for TBVS performance needed to preclude adverse effects on safety-related functions during conditions of plant operation.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.4.5, Rev 3: Engineered All Various Not Applicable This SRP Section addresses ESF ventilation Not Applicable ty Feature Ventilation systems designed for fission product em removal in a post-design basis accident environment. The NuScale design does not rely on ESF ventilation systems to mitigate the consequences of a design basis accident. Nonsafety-related normal ventilation systems provide atmosphere cleanup capability, as necessary, that meets the design, testing, and maintenance guidelines specified in RG 1.140. These systems are not credited for meeting applicable offsite dose limits. 9.5.1.1, Rev 0: Fire II.1 Fire Protection Probabilistic Risk Not Applicable Development and implementation of a risk- Not Applicable ection Program Assessment (Including Appendix C) informed, performance-based fire protection program is the responsibility of COL applicants that reference the NuScale design, and that elect to implement the provisions of 10 CFR 50.48(c). 9.5.1.1, Rev 0: Fire II.2 Fire Protection Program Not Applicable This acceptance criterion is applicable only Not Applicable ection Program Considerations for License Renewal to reactor licensees seeking license renewal. (Including Appendix B) 9.5.1.1, Rev 0: Fire II.3 NRC Staff Positions and Guidelines on Partially Conforms This acceptance criterion is applicable 9.5.1 ection Program Fire Protection except NuScale will use the current year subtier documents. Conformance with Regulatory Criteria 9.5.1.1, Rev 0: Fire II.4 Fire Protection for Permanently Not Applicable This acceptance criterion (RG 1.191) is Not Applicable ection Program Shutdown and Decommissioning applicable only to reactor licensees that Reactor Plants have submitted the necessary certifications for license termination under 10 CFR 50.82.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.5.1.1, Rev 0: Fire II.5 Fire Protection Program for New Partially Conforms This acceptance criterion and its subtier 9.5.1 ection Program Reactor Combined License guidance apply to COL applicants under Appendix 9A Applications 10 CFR 52. COL applicants referencing a certified design are responsible for implementing this guidance. Notwithstanding the above, NuScale, as an applicant for a design certification, considers this guidance to be applicable to the design certification application to the extent necessary to ensure that the COL applicant can satisfy this guidance. 9.5.1.1, Rev 0: Fire II.6 Enhanced Fire Protection Criteria for Partially Conforms The enhanced fire protection criteria for 9.5.1 ection Program New Reactor Designs (Including new reactor designs specify passive Table 9.5.1-2 Appendix A) separation of redundant trains as the Appendix 9A preferred approach to ensure safe-shutdown capability. Due to the modular nature and small size of the NuScale Power Module, it is not feasible in all instances to provide installed passive separation of redundant trains. When train separation is not feasible, fire protection for redundant shutdown systems is employed to ensure, to the extent practicable, such that one shutdown division will be free of fire damage. Conformance with Regulatory Criteria 9.5.1.1, Rev 0: Fire II.7 Operational Program and Proposed Not Applicable This acceptance criterion is the Not Applicable ection Program Implementation Milestones responsibility of the COL applicant. 9.5.1.2, Rev 0: Risk- All Various Not Applicable Development and implementation of a risk- Not Applicable med, Performance- informed, performance-based fire d Fire Protection protection program is the responsibility of ram COL applicants that reference the NuScale design, and that elect to implement the provisions of 10 CFR 50.48(c). S 9.5.2, Rev 0: II.1 Emergency Facilities and Equipment Partially Conforms This acceptance criterion is the 9.5.2 munication Systems responsibility of the COL applicant.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 9.5.2, Rev 0: II.2 Onsite Technical Support Center and Partially Conforms The NuScale standard plant design will 9.5.2 munication Systems Operational Support Center include provisions for an onsite technical support center and an onsite operational support center as specified by 10 CFR 50.34(f)(2)(xxv) and this acceptance criterion. Communication systems serving these facilities in support of emergency response are the responsibility of the COL applicant. S 9.5.2, Rev 0: II.3 Emergency Facilities and Equipment Partially Conforms The NuScale design includes provisions for 9.5.2 munication Systems for Meeting 10 CFR 52.47(a)(8) design-specific emergency facilities (i.e., pertain to design features, facilities, functions, and equipment that are technically relevant to the NuScale standard plant design), consistent with 10 CFR 50.47(a)(8) and this acceptance criterion. Communication systems and equipment serving these facilities in support of emergency response are the responsibility of the COL applicant. S 9.5.2, Rev 0: II.4 Design, Fabrication, Erection, Not Applicable None. Not Applicable munication Systems Construction, Testing, and Inspection of SSC to Meet 10 CFR 50.55a S 9.5.2, Rev 0: II.5 ITAAC Conforms The aspects of this acceptance criterion Ch 14 munication Systems within the scope of the NuScale design are Conformance with Regulatory Criteria applicable to the DCA. Aspects related to site-specific design, fabrication, erection, construction, testing, and inspection of SSC, and maintenance of records for activities throughout the life of the facility, are the responsibility of the COL applicant referencing the certified design. S 9.5.2, Rev 0: II.6 ITAAC for a COL applicant Not Applicable COL applicant responsibility to prepare Not Applicable munication Systems COL-specific ITAAC. S 9.5.2, Rev 0: II.7 Compliance with GDC 1 Conforms Site-specific scope is the responsibility of 9.5.2 munication Systems the COL applicant referencing the certified design.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 9.5.2, Rev 0: II.8 Compliance with GDC 2 Conforms None. 9.5.2 munication Systems S 9.5.2, Rev 0: II.9 Compliance with GDC 3 Conforms None. 9.5.2 munication Systems S 9.5.2, Rev 0: II.10 Compliance with GDC 4 Conforms None. 9.5.2 munication Systems S 9.5.2, Rev 0: II.11 Compliance with GDC 19 Departure The NuScale design supports an exemption 9.5.2 munication Systems from GDC 19. As described in Section 3.1.2, the design complies with a NuScale-specific principal design criterion (PDC) in lieu of this GDC. Design documents meet requirements of PDC-19 for ensuring that communication equipment is provided at appropriate locations inside the control room with the capability to support all normal and emergency operations, including intra-plant communications and plant to emergency facilities and off-site communication requirements even in the event of a single failure within a communication subsystem or the loss of the normal power source. The design addresses control room communications so that control room can maintain communications with site and offsite entities during normal Conformance with Regulatory Criteria and accident conditions. S 9.5.2, Rev 0: II.12 Compliance with Not Applicable This acceptance criterion is applicable only Not Applicable munication Systems 10 CFR 73.45(e)(2)(iii), to licensees subject to 10 CFR 73.45 and the 10 CFR 73.45(g)(4)(i), and general performance requirements of 10 CFR 73.45(g)(4)(ii) 10 CFR 73.20. The NuScale design does not reprocess spent fuel or use or transport special nuclear material.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 9.5.2, Rev 0: II.13 Compliance with 10 CFR 73.46(f) Conforms Site-specific, programmatic aspects of 9.5.2 munication Systems physical security communication systems are the responsibility of the COL applicant referencing the certified design. Aspects of this acceptance criterion related to the physical design of the power reactor and communication systems are within the scope of the certified design and are applicable to the DCA. S 9.5.2, Rev 0: II.14 Compliance with Conforms Site-specific, programmatic aspects of 13.6 (via Security munication Systems 10 CFR 73.55(e)(9)(vi)(B) physical security communication systems Technical Report) are the responsibility of the COL applicant referencing the NuScale design. Aspects of this acceptance criterion that are related to the physical design of the power reactor and communication systems within the scope of the certified design are applicable to the DCA. S 9.5.2, Rev 0: II.15 Compliance with 10 CFR 73.55(j) Partially Conforms Design focus pertains to addressing 13.6 munication Systems requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage and the communication requirements necessary for this protection. Elements of this design fall under the COL applicant and are addressed Conformance with Regulatory Criteria as part of the facility physical security plan. 9.5.3, Rev 3: Lighting II.1 Integrated Design of the System Conforms None. 9.5.3 ems 9.5.3, Rev 3: Lighting II.2 Emergency Lighting System(s) Conforms None. 9.5.3 ems 9.5.3, Rev 3: Lighting II.3 Lighting Levels Conforms None. 9.5.3 ems

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 9.5.4, Rev 3: Emergency All (II.1 thru Compliance with GDC 2, GDC 4, GDC Not Applicable The NuScale plant design does not require Not Applicable el Engine Fuel Oil Storage II.4) 5, and GDC 17 or include safety-related emergency diesel Transfer System generators that are subject to this SRP section. No AC or DC power is relied upon for the performance of NuScale plant safety functions. 9.5.5, Rev 3: Emergency All (II.1 thru Compliance with GDC 2, GDC 4, GDC Not Applicable The NuScale plant design does not require Not Applicable el Engine Cooling Water II.7) 5, GDC 17, GDC 44, GDC 45, and GDC or include safety-related emergency diesel em 46 generators that are subject to this SRP section. No AC or DC power is relied upon for the performance of NuScale plant safety functions. 9.5.6, Rev 3: Emergency All (II.1 thru Compliance with GDC 2, GDC 4, GDC Not Applicable The NuScale plant design does not require Not Applicable el Engine Starting System II.4) 5, and GDC 17 or include safety-related emergency diesel generators that are subject to this SRP section. No AC or DC power is relied upon for the performance of NuScale plant safety functions. 9.5.7, Rev 3: Emergency All (II.1 thru Compliance with GDC 2, GDC 4, GDC Not Applicable The NuScale plant design does not require Not Applicable el Engine Lubrication II.4) 5, and GDC 17 or include safety-related emergency diesel em generators that are subject to this SRP section. No AC or DC power is relied upon for the performance of NuScale plant safety functions. 9.5.8, Rev 3: Emergency All (II.1 thru Compliance with GDC 2, GDC 4, GDC Not Applicable The NuScale plant design does not require Not Applicable Conformance with Regulatory Criteria el Engine Combustion Air II.4) 5, and GDC 17 or include safety-related emergency diesel e and Exhaust System generators that are subject to this SRP section. No AC or DC power is relied upon for the performance of NuScale plant safety functions. 10.2, Rev 3: Turbine II.1 Protect SSC important to safety from Not Applicable The NuScale plant design relies on the use Not Applicable erator the effects of turbine missiles with a of barriers for the protection of SSCs turbine overspeed protection system important to safety from the effects of (GDC 4) turbine missiles.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 10.2, Rev 3: Turbine II.2 Inservice Inspection covering valves Not Applicable The NuScale plant design relies on the use Not Applicable erator essential for overspeed protection. of barriers for the protection of SSCs important to safety from the effects of turbine missiles. 10.2, Rev 3: Turbine II.3 Prevention of Adverse Effects on Not Applicable There are no safety-related SSC in the Not Applicable erator Safety-Related SSC in the Turbine Turbine Building. Building S 10.2.3, Rev 0: Turbine II.1 Materials Selection Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable r Integrity Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 do not have to rely on the turbine missile generation probabilities, including turbine rotor integrity. S 10.2.3, Rev 0: Turbine II.2 Fracture Toughness Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable r Integrity Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 do not have to rely on the turbine missile generation probabilities, including turbine rotor integrity. S 10.2.3, Rev 0: Turbine II.3 Pre-Service Inspection Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable r Integrity Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 do not have to rely on the turbine missile generation probabilities, including turbine rotor integrity. Conformance with Regulatory Criteria S 10.2.3, Rev 0: Turbine II.4 Turbine Rotor Design Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable r Integrity Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 do not have to rely on the turbine missile generation probabilities, including turbine rotor integrity. S 10.2.3, Rev 0: Turbine II.5 Inservice Inspection Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable r Integrity Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 do not have to rely on the turbine missile generation probabilities, including turbine rotor integrity.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 10.2.3, Rev 0: Turbine II.6 10 CFR 52.47(b)(1) ITAAC Not Applicable Per DSRS 10.2.3, Section I and DSRS 3.5.1.3, Not Applicable r Integrity Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 do not have to rely on the turbine missile generation probabilities, including turbine rotor integrity. S 10.3, Rev 0: Main Steam II.1 Protection against natural Conforms The NuScale main steam system (MSS) is not 10.3.1 ply System phenomena (GDC 2) safety-related, but the portion of the system downstream of the main steam isolation valves (MSIV) inside the RXB includes the secondary MSIVs which act as backup to the MSIVs. Functionality is ensured by the design and construction of the MSS to the provisions of RG 1.29, Staff Regulatory Guidance C.1.i and C.2. S 10.3, Rev 0: Main Steam II.2 Protection of SSC important to safety Conforms The NuScale MSS is not safety-related or 10.3.1 ply System from the effects of turbine missiles risk-significant. Thus, the applicability of (GDC 4) GDC 4 to the NuScale MSS reviewed under this acceptance criterion is limited to aspects ensuring that a failure of the nonsafety-related SSC does not result in an adverse effect on a safety-related SSC. S 10.3, Rev 0: Main Steam II.3 Shared SSC important to safety Conforms None. 10.3.1 ply System perform required safety functions (GDC 5) Conformance with Regulatory Criteria S 10.3, Rev 0: Main Steam II.4 MSS is capable of supporting core Partially Conforms The intent of this acceptance criterion and 10.3.1 ply System cooling or safe-shutdown (non-DBA) its subtier guidance is applicable. The in the event of an SBO (10 CFR 50.63) NuScale plant design meets the intent of this guidance with its passive design and reduced reliance on AC power to cope with design basis events. S 10.3, Rev 0: Main Steam II.5 Protection of Important-to-Safety SSC Conforms None. 10.3.1 ply System from Tornado Missiles (RG 1.117, Appendix Positions 2 and 4) 10.3.6, Rev 3: Steam and II.1 Materials Selection and Fabrication of Not Applicable The NuScale design contains no Class 2 or 3 Not Applicable water System Materials Class 2 and 3 Components components.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 10.3.6, Rev 3: Steam and II.2 Fracture Toughness of Class 2 and 3 Not Applicable The NuScale design contains no Class 2 or 3 Not Applicable water System Materials Components components. 10.4.1, Rev 3: Main II.1 Prevent excessive releases of Conforms None. 10.4.1 densers radioactivity to the environment (GDC 60) 10.4.2, Rev 3: Main II.1 Prevent excessive releases of Conforms None. 10.4.2 denser Evacuation radioactivity to the environment em (GDC 60) 10.4.3, Rev 3: Turbine II.1 Prevent excessive releases of Conforms None. 10.4.3 d Seal radioactivity to the environment (GDC 60) 10.4.4, Rev 3: Turbine II.1 Piping Failures (GDC 4) Conforms None. 10.4.4 ss System 10.4.4, Rev 3: Turbine II.2 Residual Heat Removal (GDC 34) Departure The NuScale design supports an exemption 10.4.4 ss System from the power provisions of GDC 34. As described in Section 3.1.4, the design complies with a NuScale-specific principal design criterion in lieu of this GDC. 10.4.4, Rev 3: Turbine II.3 MSIV Alternate Leakage Path Not Applicable BWR only. Not Applicable ss System 10.4.5, Rev 3: Circulating II.1 Flooding of SSC important to safety Conforms None. 10.4.5 er System (GDC 4) 10.4.6, Rev 3: Condensate II.1 Maintain direct cycle BWR plant water Not Applicable BWR only. Not Applicable nup System quality to avoid corrosion-induced Conformance with Regulatory Criteria failure of the reactor coolant pressure boundary (GDC 14) 10.4.6, Rev 3: Condensate II.2 Maintain indirect cycle PWR water Conforms In the NuScale SG design, the primary water 10.4.6 nup System quality to avoid corrosion-induced is outside the steam generator tubes, the failure of the reactor coolant pressure secondary water is inside the tubes, and boundary (GDC 14) there is no SG blowdown so the secondary chemistry requirements for the NuScale design differ from those outlined in the referenced EPRI report. S 10.4.7, Rev 0: II.1 Seismic Events (GDC 2) Conforms None. 10.4.7 densate and Feedwater em

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 10.4.7, Rev 0: II.2 Fluid Instabilities (GDC 4) Partially Conforms The intent of this acceptance criterion and 10.4.7 densate and Feedwater its subtier guidance - to satisfy GDC 4 em related to protecting SSC from fluid flow instability effects such as water hammer - is applicable. However, much of the specific language in the subtier guidance refers to reactor plant designs such as large LWRs, and is not relevant to the NuScale plant design. S 10.4.7, Rev 0: II.3 Sharing of Structures, Systems, and Conforms None. 10.4.7 densate and Feedwater Components (GDC 5) em S 10.4.7, Rev 0: II.4 Heat Removal Capability (GDC 44) Not Applicable The CFWS is not a system used to transfer Not Applicable densate and Feedwater heat to an ultimate heat sink. em S 10.4.7, Rev 0: II.5 Inspection (GDC 45) Not Applicable The CFWS is not a system used to transfer Not Applicable densate and Feedwater heat to an ultimate heat sink. em S 10.4.7, Rev 0: II.6 Testing (GDC 46) Not Applicable The CFWS is not a system used to transfer Not Applicable densate and Feedwater heat to an ultimate heat sink. em S 10.4.7, Rev 0: II.7 Flow Accelerated Corrosion Conforms None. 10.4.7 densate and Feedwater em 10.4.8, Rev 3: Steam All Various Not Applicable The NuScale steam generator design does Not Applicable Conformance with Regulatory Criteria erator Blowdown System not use a blowdown system. 10.4.9, Rev 3: Auxiliary All Various Not Applicable The NuScale design neither requires nor Not Applicable water System (PWR) uses an auxiliary feedwater system. The NuScale decay heat removal system (DHRS) performs some functions similar to an auxiliary feedwater system. However, compared to an auxiliary feedwater system, the DHRS differs in its design, operation, and relationship to the small break LOCA plant response.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 10-1, Rev 3: Design All Design Guidelines for Auxiliary Not Applicable This guidance is applicable only to large Not Applicable elines for Auxiliary Feedwater System Pump Drive and PWRs that use Auxiliary Feedwater (AFW) water System Pump Power Supply Diversity system pumps powered by electrical and e and Power Supply steam sources. The NuScale DHRS fulfills a rsity for Pressurized similar function as the AFW system at a er Reactor Plants large PWR. The NuScale DHRS design does not use pumps: it operates via passive natural circulation. 10-2, Rev 4: Design TFSGD B.1 thru Top-Feed Steam Generator Designs Not Applicable The NuScale plant design does not use a Not Applicable elines for Avoiding B.4 top-feed steam generator design. er Hammers in Steam erators 10-2, Rev 4: Design PSGD B.1 thru Preheat Steam Generator Designs Not Applicable The NuScale plant design does not use a Not Applicable elines for Avoiding B.4 preheat steam generator design. er Hammers in Steam erators 10-2, Rev 4: Design OTSGD B.1 Once-Through Steam Generator Not Applicable This acceptance criterion is applicable only Not Applicable elines for Avoiding Designs - Auxiliary Feedwater Supply to large PWRs that use a once-through er Hammers in Steam steam generator design. The NuScale plant erators design does not involve an AFW system as would be found at a typical large LWR, but does include the DHRS that fulfills a similar function as a typical AFW system. However, the NuScale steam generator design precludes potential water hammer issues Conformance with Regulatory Criteria without providing DHRS water through an externally mounted supply top discharge header as is prescribed by this acceptance criterion. 10-2, Rev 4: Design OTSGD B.2 Once-Through Steam Generator Conforms None. 5.4.1 elines for Avoiding Designs - Tests and Test Procedures er Hammers in Steam erators S 11.1, Rev 0: Coolant II.1 RG 1.110 Partially Conforms See RG 1.110 in Table 1.9-2. 11.1 ce Terms S 11.1, Rev 0: Coolant II.2 RG 1.112 Partially Conforms See RG 1.112 in Table 1.9-2. 11.1 ce Terms

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 11.1, Rev 0: Coolant II.3 RG 1.140 Partially Conforms RG 1.140 in Table 1.9-2. 11.1 ce Terms S 11.1, Rev 0: Coolant II.4 DC/COL-ISG-5 Not Applicable The NuScale design is similar to large PWRs Not Applicable ce Terms in the existing fleet for effluent release calculations. However, an alternate methodology is used because the existing PWRGALE code was developed in the 1980s for evaluation of the PWR reactors of that time and does not address the NuScale plant design. S 11.1, Rev 0: Coolant II.5 normal operation and AOO sources of Conforms None. 11.1 ce Terms radioactive liquid and gaseous effluents S 11.1, Rev 0: Coolant II.6 Release rates should be developed Partially Conforms The NuScale design is similar to large PWRs 11.1 ce Terms using methods that are consistent in the existing fleet for effluent release with NUREG-0017, PWR-GALE86, or calculations. However, an alternate ANSI/ANS 18.1-1999 methodology is used because the existing PWRGALE code was developed in the 1980s for evaluation of the large PWR reactors of that time and does not address the NuScale plant design. Some aspects of ANSI/ANS 18.1 are used for the coolant source terms. S 11.1, Rev 0: Coolant II.7 Decontamination factors used to Partially Conforms Decontamination factors are consistent 11.1 ce Terms reduce gaseous effluent releases to with the NuScale Technical Report, Effluent the environment Release (GALE Replacement) Methodology Conformance with Regulatory Criteria and Results, Revision 0 (TR-1116-52065). S 11.1, Rev 0: Coolant II.8 Decontamination factors applied to Partially Conforms Decontamination factors are consistent 11.1 ce Terms reduce liquid effluent releases to the with the NuScale Technical Report, Effluent environment Release (GALE Replacement) Methodology and Results, Revision 0 (TR-1116-52065). S 11.1, Rev 0: Coolant II.9 RWMS system augmentations used in Partially Conforms See RG 1.110 in Table 1.9-2. 11.1 ce Terms cost-benefit calculations are consistent with the guidance of RG 1.110 S 11.1, Rev 0: Coolant II.10 Primary and secondary coolant Conforms None. 11.2 ce Terms source terms, used in characterizing 11.3 liquid and gaseous effluents

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 11.1, Rev 0: Coolant II.11 If neutron activation products are Conforms None. 11.1 ce Terms expected in reactor pool water and secondary coolant S 11.1, Rev 0: Coolant II.12 10 CFR 50.34(b)(3), 10 CFR 50.34a, Partially Conforms The NuScale design is similar to large PWRs 11.1 ce Terms and 10 CFR 52.79(a)(3). in the existing fleet for effluent release calculations. However, an alternate methodology is used because the existing PWRGALE code was developed in the 1980s for evaluation of the large PWR reactors of that time and does not address the NuScale plant design. S 11.1, Rev 0: Coolant II.13 The design basis coolant source term Partially Conforms The design basis coolant source term for 11.1 ce Terms is based on a combination of NuScale is partially based on a failed fuel assumptions of failed fuel fractions fraction much less than 0.25 percent, which is described in NuScales Technical Report, Effluent Release (GALE Replacement) Methodology and Results, Revision 0 (TR-1116-52065). S 11.1, Rev 0: Coolant II.14 calculational technique or any source Conforms None. 11.1 ce Terms term parameter S 11.2, Rev 0: Liquid II.1 Capability to Meet Dose Design Partially Conforms This acceptance criterion is applicable 11.2.3 e Management System Objectives except for aspects that are related to performance of a site-specific cost-benefit analysis, which is the responsibility of the COL applicant. Conformance with Regulatory Criteria S 11.2, Rev 0: Liquid II.2 Design for Anticipated Processing Conforms None. 11.2.2 e Management System Requirements S 11.2, Rev 0: Liquid II.3 Seismic Design of Structures Housing Conforms None. 11.2.2 e Management System Liquid Waste Management System Components S 11.2, Rev 0: Liquid II.4 Provisions to Control Leakage and Conforms None. 11.2.2 e Management System Facilitate Operation and Maintenance S 11.2, Rev 0: Liquid II.5 Automatic control features Conforms None. 11.2 e Management System 11.5 11.6

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 11.2, Rev 0: Liquid II.6 Exhaust ventilation system Conforms None. 11.3 e Management System S 11.2, Rev 0: Liquid II.7 Criteria for Early Site Permit Not Applicable This acceptance criterion is applicable only Not Applicable e Management System Applications to applicants for an early site permit. S 11.3, Rev 0: Gaseous II.1 Capability to Meet Dose Design Partially Conforms This acceptance criterion is applicable 11.3.1 e Management System Objectives except for aspects that are related to 11.3.2 performance of a site-specific cost-benefit 11.3.3 analysis, which is the responsibility of the 11.3.4 COL applicant. S 11.3, Rev 0: Gaseous II.2 Design for Anticipated Processing Conforms None. 11.3.2 e Management System Requirements S 11.3, Rev 0: Gaseous II.3 Seismic Design and Quality Group Conforms None. 11.3.1 e Management System Classification of Components and Structures Housing Gaseous Waste Management System S 11.3, Rev 0: Gaseous II.4 Features to Minimize Contamination, Partially Conforms This acceptance criterion is applicable 11.3.2 e Management System Facilitate Decommissioning, and except for aspects that govern site-specific Minimize Generation of Radwaste activities that are the responsibility of the COL applicant. S 11.3, Rev 0: Gaseous II.5 Design, Testing, and Maintenance of Conforms None. 11.3.1 e Management System HEPA Filters and Charcoal Adsorbers 11.3.4 S 11.3, Rev 0: Gaseous II.6 Automatic control features Conforms None. 11.3.7 e Management System S 11.3, Rev 0: Gaseous II.7 Design to Withstand Effects of Conforms None. 11.3.2 Conformance with Regulatory Criteria e Management System Hydrogen Explosion S 11.3, Rev 0: Gaseous II.8 Postulated Leakage or Failure of a Conforms None. 11.3.3 e Management System Waste Gas Storage Tank or Offgas Charcoal Delay Bed S 11.3, Rev 0: Gaseous II.9 Criteria for Early Site Permit Not Applicable This acceptance criterion is applicable only Not Applicable e Management System Applications to applicants for an early site permit. S 11.3, Rev 0: Gaseous II.10 Relevant RGs, ISG, and BTP Partially Conforms As described above in acceptance criteria As listed above in e Management System II.1, II.3, II.4, II.5, and II.8. acceptance criteria II.1, II.3, II.4, II.5, and II.8.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 11.4, Rev 0: Solid Waste II.1 Design Parameters Based on Conforms None. Table 11.4-1 agement System Expected Radionuclide Distributions Table 11.4-5 thru and Concentrations Table 11.4-9 S 11.4, Rev 0: Solid Waste II.2 Sizing of Processing Equipment Conforms None. 11.4.2 agement System S 11.4, Rev 0: Solid Waste II.3 Liquid and Wet Waste Stabilization in Partially Conforms This acceptance criterion is applicable 11.4.2 agement System Accordance with Process Control except for aspects related to development Program and implementation of a Process Control Program (PCP), which is the responsibility of the COL applicant. S 11.4, Rev 0: Solid Waste II.4 Stabilization of Other Forms of Wet Not Applicable The development and implementation of a Not Applicable agement System Waste in Accordance with Process PCP is the responsibility of the COL Control Program applicant. S 11.4, Rev 0: Solid Waste II.5 Design Objectives, Design Criteria, Not Applicable The development and implementation of a Not Applicable agement System Treatment Methods, Expected PCP and ODCM are the responsibility of the Effluent Releases, Monitoring and COL applicant. Control Instrumentation Setpoints S 11.4, Rev 0: Solid Waste II.6 Waste Containers, Shipping Casks, Partially Conforms This guidance is applicable to design 11.4.1 agement System and Waste Packaging certification except for site-specific, 11.4.2 programmatic and operational aspects that are the responsibility of the COL applicant. S 11.4, Rev 0: Solid Waste II.7 Onsite Waste Storage Facilities Partially Conforms This guidance is applicable to design 11.4.1 agement System certification except for site-specific, 11.4.2 programmatic and operational aspects that are the responsibility of the COL applicant. Conformance with Regulatory Criteria S 11.4, Rev 0: Solid Waste II.8 Seismic Design and Quality Group Conforms None. 3.8 agement System Classification of Components and 11.4.1 Structures Housing Solid Waste Table 11.4-1 Management System 11.4.2 S 11.4, Rev 0: Solid Waste II.9 Provisions to Control Leakage and Partially Conforms This acceptance criterion is applicable 11.4.1 agement System Facilitate Operation and Maintenance except for aspects that govern site-specific 11.4.3 activities that are the responsibility of the 12.3 COL applicant.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 11.4, Rev 0: Solid Waste II.10 Features to Minimize Contamination, Partially Conforms This acceptance criterion is applicable 11.4.1 agement System Facilitate Decommissioning, and except for aspects that govern site-specific 11.4.3 Minimize Generation of Radwaste activities that are the responsibility of the 12.3 COL applicant. S 11.4, Rev 0: Solid Waste II.11 Storage Facility Design for Long Term Not Applicable The NuScale design has no long term Not Applicable agement System Onsite Storage (Including Appendix storage facility for solid radioactive waste. 11.4A) This is a COL applicant responsibility. S 11.4, Rev 0: Solid Waste II.12 Class A, B, C - Processing and Partially Conforms This acceptance criterion governs site- 11.4.2 agement System Disposing of Liquid, Wet, and Dry specific, programmatic aspects of the PCP 11.4.3 Solid Wastes development and implementation that are the responsibility of the COL applicant. This guidance is applicable to the extent necessary to ensure that the COL applicant referencing the certified design can satisfy the guidance. S 11.4, Rev 0: Solid Waste II.13 Greater than Class C - Processing and Partially Conforms This acceptance criterion governs site- 11.4.1 agement System Disposing of Liquid, Wet, and Dry specific, programmatic aspects of the PCP 11.4.2 Solid Wastes development and implementation that are the responsibility of the COL applicant. This guidance is applicable to the extent necessary to ensure that the COL applicant referencing the certified design can satisfy the guidance. S 11.4, Rev 0: Solid Waste II.14 Processing and Disposing of Mixed Partially Conforms This acceptance criterion governs site- 11.4.2 agement System Wastes specific, programmatic aspects of PCP Conformance with Regulatory Criteria implementation (specific to mixed waste processing and disposal) that are the responsibility of the COL applicant. This guidance is applicable to the extent necessary to ensure that the COL applicant referencing the certified design can satisfy the guidance contained therein. S 11.4, Rev 0: Solid Waste II.15 All Effluent Releases Associated with Partially Conforms This acceptance criterion is applicable 11.4.2 agement System Operation of the SWMS except for site specific, programmatic aspects that are the responsibility of the COL applicant.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 11.4, Rev 0: Solid Waste II.16 Operational Programs Not Applicable The information governed by this Not Applicable agement System acceptance criterion is site-specific and is the responsibility of the COL applicant. S 11.4, Rev 0: Solid Waste II.17 Automatic control features Not Applicable Not Applicable agement System S 11.4, Rev 0: Solid Waste II.18 Design of exhaust ventilation systems Conforms None. 11.4.2 agement System S 11.4, Rev 0: Solid Waste II.19 Seismic design of structures housing Conforms None. 11.4.1 agement System SWMS S 11.5, Rev 0: Process and II.1 Installation of Instrumentation and Partially Conforms This acceptance criterion is applicable 9.3.2 ent Radiological Monitoring Equipment and Sampling except for certain aspects of its subtier 11.2 itoring Instrumentation and Analyses of Normal and Potential guidance (see RG 1.21, RG 1.33, RG 1.97, RG 11.3 Sampling Systems Effluent Pathways 4.1, RG 4.15, and BTP 7-10). 11.5 S 11.5, Rev 0: Process and II.2 Instrumentation and Monitoring Partially Conforms This acceptance criterion is applicable 9.3.2 ent Radiological Equipment and Sampling and except for certain aspects of its subtier 11.5 itoring Instrumentation Analysis of Radioactive Waste Process guidance (see RG 1.21, RG 1.33, RG 1.97, RG 12.3.4 Sampling Systems Systems (Including Appendix 11.5A) 4.15, RG 4.21 and BTP 7-10). Administrative and procedural controls are COL applicant responsibility. S 11.5, Rev 0: Process and II.3 Provisions for Administrative and Partially Conforms This acceptance criterion is applicable 9.3.2 ent Radiological Procedural Controls (Including except for certain aspects of its subtier 11.5 itoring Instrumentation Appendix 11.5A) guidance (see RG 1.21, RG 1.33, RG 1.97 and 12.3 Sampling Systems RG 4.15). Administrative and procedural controls are COL applicant responsibility. Conformance with Regulatory Criteria S 11.5, Rev 0: Process and II.4 Monitoring, Sampling, and Analyses Partially Conforms This acceptance criterion is applicable 11.5 ent Radiological of All Identified Gaseous Effluent except for certain aspects of its subtier itoring Instrumentation Release Paths (Including Appendix guidance (see RG 1.97 and BTP 7-10). Sampling Systems 11.5A) Administrative and procedural controls are COL applicant responsibility. S 11.5, Rev 0: Process and II.5 Monitoring, Sampling, and Analysis of Partially Conforms This acceptance criterion is applicable 11.5 ent Radiological All Identified Liquid Effluent Release except for the administrative and itoring Instrumentation Paths procedural controls that are the COL Sampling Systems applicant's responsibility.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 11.5, Rev 0: Process and II.6 Operational Programs Not Applicable The information governed by this Not Applicable ent Radiological acceptance criterion is site-specific and is itoring Instrumentation the responsibility of the COL applicant. Sampling Systems S 11.5, Rev 0: Process and II.7 Descriptions of design features and Conforms None. 11.5 ent Radiological instrumentation used in primary and itoring Instrumentation secondary coolant system leakage Sampling Systems detection S 11.6, Rev 0: Guidance II.1 Installation of instrumentation or Conforms None. 11.5 strumentation and sampling equipment 11.6 rol Design Features for ess and Effluent ological Monitoring, and Radiation and Airborne oactivity Monitoring S 11.6, Rev 0: Guidance II.2 Gaseous and liquid release points Conforms None. 11.5 strumentation and should be monitored 11.6 rol Design Features for ess and Effluent ological Monitoring, and Radiation and Airborne oactivity Monitoring S 11.6, Rev 0: Guidance II.3 Radiation exposure rates and Conforms None. 11.6 strumentation and airborne concentration monitoring 12.3 rol Design Features for locations and sampling points Conformance with Regulatory Criteria ess and Effluent ological Monitoring, and Radiation and Airborne oactivity Monitoring S 11.6, Rev 0: Guidance II.4 Compliance with GDC 63 & 64 via Partially Conforms This acceptance criterion is applicable 9.3.2 strumentation and post-TMI action plan items except for aspects of its subtier regulation 11.5 rol Design Features for 10 CFR 50.34(f)(2)(xxvi) that address testing 11.6 ess and Effluent and operational programs, which are a COL ological Monitoring, and applicant responsibility. Radiation and Airborne oactivity Monitoring

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 11.6, Rev 0: Guidance II.5 Ensure samples are representative Conforms None. 9.3.2 strumentation and 11.6 rol Design Features for ess and Effluent ological Monitoring, and Radiation and Airborne oactivity Monitoring S 11.6, Rev 0: Guidance II.6 Describe process used to develop, Partially Conforms This acceptance criterion is applicable 11.6 strumentation and review, verify, validate and audit except for site-specific, programmatic Ch 17 rol Design Features for digital computer software. aspects regarding software reviews, which ess and Effluent are the COL applicant's responsibility. ological Monitoring, and Radiation and Airborne oactivity Monitoring S 11.6, Rev 0: Guidance II.7 RETS/SREC and ODCM established Not Applicable The RETS/SREC and ODCM are COL Not Applicable strumentation and setpoints. applicant responsibilities. rol Design Features for ess and Effluent ological Monitoring, and Radiation and Airborne oactivity Monitoring S 11.6, Rev 0: Guidance II.8 Compliance with 10 CFR 20.1406 via Partially Conforms See RG 4.21 in Table 1.9-2. 11.5 strumentation and RG 4.21, NEI 97-06, 08-08A and 07-07. 11.6 rol Design Features for 12.3.6 ess and Effluent Conformance with Regulatory Criteria ological Monitoring, and Radiation and Airborne oactivity Monitoring S 11.6, Rev 0: Guidance II.9 Description of design features and Conforms None. 5.2.5 strumentation and instrumentation used in primary and 9.3.4 rol Design Features for secondary coolant system leakage 9.3.6 ess and Effluent detection 11.5 ological Monitoring, and 11.6 Radiation and Airborne oactivity Monitoring

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 11.6, Rev 0: Guidance II.10 Additional information on operating Conforms None. 11.5 strumentation and experience 11.6 rol Design Features for 12.3 ess and Effluent ological Monitoring, and Radiation and Airborne oactivity Monitoring S 11.6, Rev 0: Guidance II.11 Radiation monitoring and sampling Conforms None. 13.4 strumentation and conformance to Tech Specs, Initial 14.2 rol Design Features for Test Program, and ITAAC. 14.3 ess and Effluent Ch 16 ological Monitoring, and Radiation and Airborne oactivity Monitoring S 11.6, Rev 0: Guidance II.12 Describe the types and ranges of Conforms None. 11.5 strumentation and radiation monitoring equipment 11.6 rol Design Features for 12.3 ess and Effluent ological Monitoring, and Radiation and Airborne oactivity Monitoring S 11.6, Rev 0: Guidance II.13 Reactor fuel storage area monitors Conforms None. 7.2 strumentation and 11.5 rol Design Features for 11.6 ess and Effluent 12.3.4 Conformance with Regulatory Criteria ological Monitoring, and Radiation and Airborne oactivity Monitoring 11-3, Rev 4: Design B.1 Processing Requirements Conforms This guidance is applicable except for 11.4.1 ance for Solid aspects related to PCP development and 11.4.2 oactive Waste implementation that are applicable to COL agement Systems applicants. lled in Light-Water-ed Nuclear Power tor Plants

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 11-3, Rev 4: Design B.2 Assurance of Complete Stabilization Not Applicable This guidance is related to PCP Not Applicable ance for Solid or Dewatering development and implementation that are oactive Waste applicable to COL applicants. agement Systems lled in Light-Water-ed Nuclear Power tor Plants 11-3, Rev 4: Design B.3 Waste Storage Conforms None. 11.4.1 ance for Solid 11.4.2 oactive Waste agement Systems lled in Light-Water-ed Nuclear Power tor Plants 11-3, Rev 4: Design B.4 Portable Solid Waste Systems Partially Conforms This guidance is applicable except for 11.4.1 ance for Solid aspects related to control and use of 11.4.2 oactive Waste portable solid radwaste processing agement Systems equipment that are applicable to COL lled in Light-Water- applicants. ed Nuclear Power tor Plants 11-3, Rev 4: Design B.5 Additional Design Features Partially Conforms This guidance is applicable except for 11.4.2 ance for Solid aspects related to PCP development and oactive Waste implementation that are applicable to COL agement Systems applicants. Conformance with Regulatory Criteria lled in Light-Water-ed Nuclear Power tor Plants 11-5, Rev 4: Postulated B.1 Waste Gas System Leak or Failure Partially Conforms This acceptance criterion is applicable 11.3.1 oactive Releases Due to a Analysis except for aspects that are BWR-specific or 11.3.3 e Gas System Leak or are site-specific. re 11-5, Rev 4: Postulated B.2 Staff Method for Analysis Conforms None. 11.3.3 oactive Releases Due to a e Gas System Leak or re

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 11-6, Rev 4: Postulated B.1 Failure Mechanism and Radioactivity Not Applicable COL applicant. Not Applicable oactive Releases Due to Releases id-Containing Tank res 11-6, Rev. 4: Postulated B.2 Mitigating Design Features Not Applicable COL applicant. Not Applicable oactive Releases Due to id-Containing Tank res 11-6, Rev 4: Postulated B.3 Radioactive Source Term Partially conforms This acceptance criterion is applicable 11.2.3 oactive Releases Due to except for aspects that are BWR-specific or id-Containing Tank are related to site-specific activities that are res the responsibility of the COL applicant. 11-6, Rev 4: Postulated B.4 Calculations of Transport Capabilities Not Applicable The development of representative site Not Applicable oactive Releases Due to in Groundwater or Surface Water parameters under this acceptance criterion id-Containing Tank (and SRP Section 2.4.13) is site-specific and res applicable to COL applicant. 11-6, Rev 4: Postulated B.5 Exposure Scenarios and Acceptance Not Applicable The development of representative site Not Applicable oactive Releases Due to Criteria parameters under this acceptance criterion id-Containing Tank is site-specific and applicable to COL res applicant. 11-6, Rev 4: Postulated B.6 SRP Dose Acceptance Criteria Not Applicable This acceptance criterion is the Not Applicable oactive Releases Due to responsibility of the COL applicant. id-Containing Tank res 11-6, (Rev 4): Postulated B.7 Specifications on Tank Waste Not Applicable Compliance with this guidance is the Not Applicable Conformance with Regulatory Criteria oactive Releases Due to Radioactivity Concentration Levels responsibility of the COL applicant. id-Containing Tank res 12.1, Rev 4: Assuring That II.1 Policy Considerations Partially Conforms These site-specific aspects are the 12.1.1 pational Radiation responsibility of the COL applicant sures Are As Low As Is referencing the certified design. onably Achievable 12.1, Rev 4: Assuring That II.2 Design Considerations Conforms None. 12.1.2 pational Radiation sures Are As Low As Is onably Achievable

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 12.1, Rev 4: Assuring That II.3 Operational Considerations Not Applicable This guidance governs site-specific Not Applicable pational Radiation operational programs, plans, and sures Are As Low As Is procedures that are the responsibility of the onably Achievable COL applicant. 12.1, Rev 4: Assuring That II.4 Radiation Protection Considerations Not Applicable See comment above for Acceptance Not Applicable pational Radiation Criterion II.3. sures Are As Low As Is onably Achievable S 12.2, Rev 0: Radiation II.1 RG 1.183 Partially Conforms See RG 1.183 in Table 1.9-2. 12.2.1 ces S 12.2, Rev 0: Radiation II.2 RG 1.7 Not Applicable See RG 1.7 in Table 1.9-2. Not Applicable ces There is no radiation source created from the determination of gaseous concentrations in containment following an accident (such as sample lines outside containment). S 12.2, Rev 0: Radiation II.3 RG 1.112 Partially Conforms See RG 1.112 in Table 1.9-2. 12.2.1 ces S 12.2, Rev 0: Radiation II.4 NUREG-0737, Task Action Plan Item Conforms None. 12.3 ces II.B.2 12.4 S 12.2, Rev 0: Radiation II.5 ANSI/ANS Standard 18.1 Conforms None. 11.1 ces S 12.2, Rev 0: Radiation II.6 Radiation Sources for 10 CFR 50.49 Conforms None. 12.2 ces (EQ) Ch 3 Conformance with Regulatory Criteria S 12.2, Rev 0: Radiation II.7 RG 1.143 Partially Conforms See RG 1.143 in Table 1.9-2. 11.2 ces 11.3 11.4 11.6 S 12.2, Rev 0: Radiation II.8 RG 1.26, RG 1.29 and RG 1.117 Conforms None. 3.2 ces S 12.3-12.4, Rev 0: II.1 RG 1.7 Not Applicable See RG 1.7 in Table 1.9-2. Not Applicable ation Protection Design There is no radiation field created from the ures determination of gaseous concentrations in containment following an accident (such as sample lines outside containment).

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 12.3-12.4, Rev 0: II.2 RG 1.52 Not Applicable See RG 1.52 in Table 1.9-2. Not Applicable ation Protection Design ures S 12.3-12.4, Rev 0: II.3 RG 1.69 Partially Conforms See RG 1.69 in Table 1.9-2. 12.3.2 ation Protection Design ures S 12.3-12.4, Rev 0: II.4 RG 1.97 Partially Conforms See RG 1.97 in Table 1.9-2. 7.2.13 ation Protection Design 12.3.4 ures S 12.3-12.4, Rev 0: II.5 RG 1.183 Partially Conforms See RG 1.183 in Table 1.9-2. 12.2 ation Protection Design ures S 12.3-12.4, Rev 0: II.6 RG 8.2 Not Applicable See RG 8.2 in Table 1.9-2. Not Applicable ation Protection Design ures S 12.3-12.4, Rev 0: II.7 RG 8.8 Partially Conforms See RG 8.8 in Table 1.9-2. 12.3.1 ation Protection Design ures S 12.3-12.4, Rev 0: II.8 RG 8.10 Not Applicable See RG 8.10 in Table 1.9-2. Not Applicable ation Protection Design ures S 12.3-12.4, Rev 0: II.9 RG 8.15 Not Applicable See RG 8.15 in Table 1.9-2. Not Applicable ation Protection Design ures Conformance with Regulatory Criteria S 12.3-12.4, Rev 0: II.10 RG 8.19 Conforms None. 12.4 ation Protection Design ures S 12.3-12.4, Rev 0: II.11 RG 8.25 Not Applicable See RG 8.25 in Table 1.9-2. Not Applicable ation Protection Design ures S 12.3-12.4, Rev 0: II.12 RG 8.38 Partially Conforms See RG 8.38 in Table 1.9-2. 12.3.1 ation Protection Design ures S 12.3-12.4, Rev 0: II.13 ANSI/ANS/HPSSC-6.8.1-1981 Conforms None. 12.3.4 ation Protection Design ures

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 12.3-12.4, Rev 0: II.14 ANSI/HPS N13.1-2011 Conforms None. 12.3.4 ation Protection Design ures S 12.3-12.4, Rev 0: II.15 ANSI/ANS-6.4-2006 Conforms None. 12.3.2 ation Protection Design ures S 12.3-12.4, Rev 0: II.16 Memo from Larry W. Camper to David Partially Conforms The portion of this guidance that pertains to 12.3.4 ation Protection Design B. Matthews and Elmo E. Collins dated the design phase is applicable to the DCA. ures 10-10-2006 S 12.3-12.4, Rev 0: II.17 RG 1.140 Partially Conforms See RG 1.140 in Table 1.9-2. 12.3.3 ation Protection Design ures S 12.3-12.4, Rev 0: II.18 RG 1.89 Partially Conforms See RG 1.89 in Table 1.9-2. 3.11 ation Protection Design ures S 12.3-12.4, Rev 0: II.19 RG 4.21 Partially Conforms See RG 4.21 in Table 1.9-2. 12.3.6 ation Protection Design ures S 12.3-12.4, Radiation II.20 RG 1.45 Partially Conforms See RG 1.45 in Table 1.9-2. 5.2.5 ection Design Features S 12.3-12.4, Rev 0: II.21 NEI 97-06 Conforms None. Ch 5 ation Protection Design ures S 12.3-12.4, Rev 0: II.22 RG 1.143 Partially Conforms See RG 1.143 in Table 1.9-2. Ch 11 Conformance with Regulatory Criteria ation Protection Design ures S 12.3-12.4, Rev 0: II.23 BTP 11-3 and SECY-94-198 Partially Conforms These guidance documents are not 11.4 ation Protection Design applicable to the DCA so far as they address ures the addition of supplemental extended LLW storage and the development of a PCP. This is a COL applicant responsibility. S 12.3-12.4, Rev 0: II.24 RG 1.97 Partially Conforms See RG 1.97 in Table 1.9-2. 12.3.4 ation Protection Design ures

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 12.3-12.4, Rev 0: II.25 RG 1.12 Partially Conforms See RG 1.12 in Table 1.9-2. 12.3.1 ation Protection Design ures S 12.5, Rev 0: Operational All Various Not Applicable This guidance governs operational Not Applicable ation Protection Program programs, procedures, facilities and organization that are site-specific, and are the responsibility of the COL applicant referencing the certified design. 13.1.1, Rev 5: All General and Specific Requirements Not Applicable COL applicant. Not Applicable agement and Technical ort Organization 13.1.2 - 13.1.3, Rev 6: All Operating Organization Not Applicable COL applicant. Not Applicable rating Organization 13.2.1, Rev 3: Reactor All General and Specific Requirements Not Applicable COL applicant. Not Applicable rator Requalification ram; Reactor Operator ing 13.2.2, Rev 3: Non- All Various Not Applicable COL applicant. Not Applicable nsed Plant Staff Training 13.3, Rev 3: Emergency II.1 Meeting the Standards of Not Applicable COL applicant. Not Applicable ning 10 CFR 50.47(b); Conduct of Full Participation Exercise per 10 CFR 50, Appendix E 13.3, Rev 3: Emergency II.2 Onsite and Offsite Emergency Not Applicable COL applicant. Not Applicable Conformance with Regulatory Criteria ning Response Plans 13.3, Rev 3: Emergency II.3 Emergency Classification and Action Not Applicable COL applicant. Not Applicable ning Level Scheme 13.3, Rev 3: Emergency II.4 Meteorological Criteria Not Applicable COL applicant. Not Applicable ning 13.3, Rev 3: Emergency II.5 Upgrading Emergency Response Not Applicable There are no proposed changes to existing Not Applicable ning Facilities emergency response facilities. 13.3, Rev 3: Emergency II.6 Alerting and Notifications Not Applicable COL applicant. Not Applicable ning 13.3, Rev 3: Emergency II.7 Protective Action Recommendations Not Applicable COL applicant. Not Applicable ning

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 13.3, Rev 3: Emergency II.8 Alternatives to NUREG-0654/FEMA- Not Applicable COL applicant. Not Applicable ning REP-1, Rev 1, 13.3, Rev 3: Emergency II.9 State, Tribal, and Local Government Not Applicable COL applicant. Not Applicable ning Planning and Preparedness 13.3, Rev 3: Emergency II.10 Emergency Planning Zones Not Applicable COL applicant. Not Applicable ning 13.3, Rev 3: Emergency II.11 Evacuation Time Estimates Not Applicable COL applicant. Not Applicable ning 13.3, Rev 3: Emergency II.12 Emergency Response Data System Partially Conforms The NuScale design includes an emergency 13.3 ning response data system. Site-specific aspects are the responsibility of the COL applicant that references the NuScale certified design. 13.3, Rev 3: Emergency II.13 Acceptability of Emergency Plans Not Applicable COL applicant. Not Applicable ning 13.3, Rev 3: Emergency II.14 Offsite Emergency Planning When Not Applicable COL applicant. Not Applicable ning Local Governments Decline to Participate 13.3, Rev 3: Emergency II.15 Early Site Permit Criteria - Physical Not Applicable Guidance applies to early site permit Not Applicable ning Characteristics Unique to Proposed applicants. Site 13.3, Rev 3: Emergency II.16 Early Site Permit Criteria - Preliminary Not Applicable Guidance applies to early site permit Not Applicable ning Analysis of Evacuation Times applicants. 13.3, Rev 3: Emergency II.17 Physical Characteristics Unique to Not Applicable Guidance applies to early site permit Not Applicable ning Proposed Site applicants. Conformance with Regulatory Criteria 13.3, Rev 3: Emergency II.18 Copies of Letters of Agreement or Not Applicable Guidance applies to early site permit Not Applicable ning Other Certifications applicants. 13.3, Rev 3: Emergency II.19 Emergency Preparedness Information Not Applicable Guidance applies to early site permit Not Applicable ning and Plans Associated with Early Site applicants. Permit Application 13.3, Rev 3: Emergency II.20 Complete and Integrated Emergency Not Applicable Guidance applies to early site permit Not Applicable ning Plans Associated with Early Site applicants. Permit Application 13.3, Rev 3: Emergency II.21 ITAAC Associated with Early Site Not Applicable Guidance applies to early site permit Not Applicable ning Permit Application applicants. 13.3, Rev 3: Emergency II.22 ITAAC Associated with Design Not Applicable Emergency planning ITAAC are not part of Not Applicable ning Certification Application the NuScale DCA.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 13.3, Rev 3: Emergency II.23 ITAAC Associated with Combined Not Applicable COL applicant. Not Applicable ning License Application 13.3, Rev 3: Emergency II.24 Generic Emergency Planning ITAAC Not Applicable COL applicant. Not Applicable ning 13.3, Rev 3: Emergency II.25 Design and Implementation of Partially Conforms The NuScale design includes a technical 13.3 ning Emergency Response Facilities support center. The operational support center and the emergency operations facility are the responsibility of the COL applicant that references the NuScale certified design. 13.3, Rev 3: Emergency II.26 Safety Parameter Display System Conforms Safety parameter displays are provided in 13.3 ning the technical support center. The emergency operations facility is the responsibility of the COL applicant that references the NuScale design certification. 13.3, Rev 3: Emergency II.27 Reactor Coolant System and Departure The NuScale design supports an exemption 9.3.2 ning Containment Sampling from 10 CFR 50.34(f)(2)(viii). 13.3, Rev 3: Emergency II.28 Containment Monitoring and Partially Conforms Programmatic aspects of containment and 9.3.2, 11.5 ning Continuous Sampling from Potential effluent monitoring are the responsibility of Accident Release Points the COL applicant. 13.3, Rev 3: Emergency II.29 NRC Notifications and Not Applicable COL applicant. Not Applicable ning Communications 13.3, Rev 3: Emergency II.30 Generic Communications and Not Applicable COL applicant. Not Applicable ning Commission Orders Pertaining to Conformance with Regulatory Criteria Emergency Planning 13.3, Rev 3: Emergency II.31 Operational Programs Not Applicable COL applicant. Not Applicable ning 13.4, Rev 3: Operational Not Applicable Various (Including Attachment, Not Applicable There are no specific requirements for this Not Applicable rams Sample FSAR Table 13.4-x) SRP section. 13.5.1.1, Rev 1: All Various Not Applicable COL applicant. Not Applicable inistrative Procedures - eral 13.5.1.2, Draft Rev 0: All Various Not Applicable Draft SRP section was never finalized. Not Applicable inistrative Procedures - Content was subsumed into SRP Section l Test Program 14.2.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 13.5.2.1, Rev 2: Operating All Various Not Applicable COL applicant. Not Applicable Emergency Operating edures 13.5.2.2, Draft Rev 0: All Various Not Applicable NRC never finalized guidance in this SRP. Not Applicable tenance and Other Instead, applicable guidance was relocated rating Procedures to SRP 17.5. 13.6.1, Rev 1: Physical All Various Not Applicable COL applicant. Not Applicable rity - Combined License Operating Reactors 13.6.2, Rev 2: Physical All Various Conforms Applicable for the physical security 13.6.2 (via Security rity - Design Certification elements within the certified design Technical Report) boundary of the NuScale plant. 13.6.3, Rev 1: Physical All Various Not Applicable ESP applicant. Not Applicable rity - Early Site Permit 13.6.4, Rev 1: Access II (no number) 10 CFR 73.56 Not Applicable COL applicant. Not Applicable orization 13.6.6, Rev 0: Cyber All Various Not Applicable COL applicant. Not Applicable rity Plan 13.7.1, Rev 0: Fitness for All Various Not Applicable COL applicant. Not Applicable (Operational) 13.7.2, Rev 0: Fitness for All Various Not Applicable COL applicant. Not Applicable (Construction) S 14.2, Rev 0: Initial Plant II.1 Summary of Test Program and Conforms None. 14.2 Program - Design Objectives Conformance with Regulatory Criteria fication and New COL icants S 14.2, Rev 0: Initial Plant II.2 Test Programs Conformance with Conforms None. 14.2 Program - Design Regulatory Guides fication and New COL icants S 14.2, Rev 0: Initial Plant II.3 Initial Test Program Administrative Partially Conforms Subheading DC Applicant, Items A through 14.2 Program - Design Procedures D, are applicable to the DCA. Subheading fication and New COL COL/OL applicants, Items A through H, are icants applicable only to COL applicant.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 14.2, Rev 0: Initial Plant II.4 Initial Startup Tests Partially Conforms Subheading DC Applicant, Item A, is 14.2 Program - Design applicable to the DCA. Subheading COL/OL fication and New COL applicants, Items A and B, are applicable to icants COL applicants. S 14.2, Rev 0: Initial Plant II.5 Individual Test Descriptions/ Conforms None. 14.2 Program - Design Abstracts fication and New COL icants S 14.2, Rev 0: Initial Plant II.6 Initial Test Program Acceptance Partially Conforms Subheading DC Applicant, Items A through 14.2 Program - Design Criteria C, are applicable to the DCA. Subheading fication and New COL COL/OL applicants, Items A through C, are icants applicable to COL applicants. 14.2.1, (August 2006): All Various Not Applicable This SRP section is applicable only to Not Applicable eric Guidelines for extended power uprate license amendment nded Power Uprate requests. ng Programs 14.3, (March 2007): II.1 Acceptability of the Scope of ITAAC Partially Conforms A portion of this acceptance criterion is 14.3 ections, Tests, Analyses, applicable to COL applicants. Acceptance Criteria 14.3, (March 2007): II.2 Specific Acceptance Criteria for ITAAC Conforms None. 14.3 ections, Tests, Analyses, Specified in SRP Section 14.3 Acceptance Criteria 14.3.2, Rev 0: Structural All Various Not Applicable Methodology for developing ITAAC is Not Applicable Systems Engineering - provided in SRP 14.3. ections, Tests, Analyses, Conformance with Regulatory Criteria Acceptance Criteria 14.3.3, (March 2007): All Various Not Applicable Methodology for developing ITAAC is Not Applicable g Systems and provided in SRP 14.3. ponents - Inspections, s, Analyses, and ptance Criteria 14.3.4, Rev 0: Reactor All Various Not Applicable Methodology for developing ITAAC is Not Applicable ems - Inspections, Tests, provided in SRP 14.3. yses, and Acceptance ria

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 14.3.5, Rev 0: All Various Not Applicable Methodology for developing ITAAC is Not Applicable umentation and Controls provided in SRP 14.3. pections, Tests, Analyses, Acceptance Criteria 14.3.6, Rev 0: Electrical All Various Not Applicable Methodology for developing ITAAC is Not Applicable ems - Inspections, Tests, provided in SRP 14.3. yses, and Acceptance ria 14.3.7, Rev 0: Plant All Various Not Applicable Methodology for developing ITAAC is Not Applicable ems - Inspections, Tests, provided in SRP 14.3. yses, and Acceptance ria 14.3.8, Rev 0: Radiation All Various Not Applicable Methodology for developing ITAAC is Not Applicable ection - Inspections, provided in SRP 14.3. s, Analyses, and ptance Criteria 14.3.9, (March 2007): All Various Not Applicable Methodology for developing ITAAC is Not Applicable an Factors Engineering - provided in SRP 14.3. ections, Tests, Analyses, Acceptance Criteria 14.3.10, (March 2007): All Various Not Applicable Methodology for developing ITAAC is Not Applicable rgency Planning - provided in SRP 14.3. ections, Tests, Analyses, Acceptance Criteria 14.3.11, (March 2007): All Various Not Applicable Methodology for developing ITAAC is Not Applicable Conformance with Regulatory Criteria ainment Systems - provided in SRP 14.3. ections, Tests, Analyses, Acceptance Criteria 14.3.12, Rev 1: Physical All Various Partially Conforms The COL applicant addresses Physical 14.3.12 rity Hardware - Security Hardware ITAAC outside of the ections, Tests, Analyses, nuclear island and structures. Acceptance Criteria S 15.0, Rev 0: Introduction I.1 Categorization of Transients and Conforms None. 15.0 nsient and Accident Accidents yses

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.0, Rev 0: Introduction I.2 Categorization According to Partially Conforms Events that have been historically classified 15.0 nsient and Accident Frequency of Occurrence as AOOs are not analyzed for frequency of yses occurrence. Some events that have an IE frequency are also deterministically classified as AOOs. S 15.0, Rev 0: Introduction I.3 Categorization According to Type Conforms None. 15.0 nsient and Accident yses S 15.0, Rev 0: Introduction I.4.A Analysis Acceptance Criteria for AOOs Conforms None. 15.0.0 nsient and Accident yses S 15.0, Rev 0: Introduction I.4.B Analysis Acceptance Criteria for IEs Partially Conforms The guidance is applicable except for 4.B.ii 15.0.0 nsient and Accident and Postulated Accidents and 4.B.iv. CHF, not DNBR, is used to yses determine the thermal margin for the fuel cladding. LOCA acceptance criteria uses an acceptance criterion that is more restrictive than the temperature limit of 2,200 degrees F. S 15.0, Rev 0: Introduction I.5 Plant Characteristics Considered in Conforms None. 15.0 nsient and Accident the Safety Evaluation yses S 15.0, Rev 0: Introduction I.6 Assumed Protection and Safety Conforms None. 15.0 nsient and Accident Systems Actions yses S 15.0, Rev 0: Introduction I.7 Evaluation of Individual Initiating Conforms None. 15.0 Conformance with Regulatory Criteria nsient and Accident Events yses S 15.0, Rev 0: Introduction I.8.A Identification of Causes and Conforms None. 15.0 nsient and Accident Frequency Classification yses S 15.0, Rev 0: Introduction I.8.B Sequence of Events and Systems Partially Conforms This acceptance criterion is applicable 15.0 nsient and Accident Operation except for Item B.vi, which is applicable to yses COL applicants.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.0, Rev 0: Introduction I.8.C Core, System, and Barrier Partially Conforms The guidance is applicable except for 15.0.2 nsient and Accident Performance aspects that are BWR-specific. NuScale yses evaluates critical heat flux (CHF), which is more applicable to the NuScale design than DNBR. 15.0.1, Rev 0: Radiological II (No number) First full paragraph and 6 bullets on Not Applicable The NuScale design uses a modified version Not Applicable equence Analyses Using Page 15.0.1-6, Compliance with of the alternative source term (AST) native Source Terms Specific Provisions of NUREG-0737 methodology to evaluate radiological consequences of accidents. 15.0.1, Rev 0: Radiological II (No number) Last paragraph on Page 15.0.1-6 and Not Applicable The NuScale design utilizes a modified Not Applicable equence Analyses Using Table 1, Exposure Criteria for version of the AST methodology to evaluate native Source Terms Radiological Consequences of Design radiological consequences of accidents. Basis Accident 15.0.2, (March 2007): II.1 Evaluation Model Departure The NuScale design supports an exemption 15.0.2 ew of Transient and from selected portions of 10 CFR 50 dent Analysis Methods Appendix K. NuScale ECCS evaluation models for LOCAs only address technically relevant features required by Appendix K. 15.0.2, (March 2007): II.2 Accident Scenario Identification Conforms None. 15.0.2 ew of Transient and Process dent Analysis Methods 15.0.2, (March 2007): II.3 Code Assessment Departure The NuScale design supports an exemption 15.0.2 ew of Transient and from selected portions of 10 CFR 50 dent Analysis Methods Appendix K. NuScale ECCS evaluation models for LOCAs only assess the Conformance with Regulatory Criteria technically relevant features required by Appendix K and TMI Action Item II.K3.30. 15.0.2, (March 2007): II.4 Uncertainty Analysis Conforms Non-LOCA methods use sensitivity analyses 15.0.2 ew of Transient and or bounding values to determine input dent Analysis Methods parameters. 15.0.2, (March 2007): II.5 Quality Assurance Plan Conforms None. 15.0.2 ew of Transient and dent Analysis Methods

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.0.3, Rev 0: Design II.1 Offsite Radiological Consequences of Conforms None. 15.0.3 Accident Radiological Postulated DBAs (includes Table 1) equence Analyses for NuScale SMR Design S 15.0.3, Rev 0: Design II.2 Control Room Radiological Conforms None. 6.4 Accident Radiological Habitability 9.4 equence Analyses for 13.3 NuScale SMR Design 15.0.3 S 15.0.3, Rev 0: Design II.3 Technical Support Center Partially Conforms Dose acceptance criterion met for TSC 15.0.3 Accident Radiological Radiological Habitability when AC power is available. TSC function is equence Analyses for transferred to the main control room when NuScale SMR Design AC is not available. S 15.1.1-15.1.4, Rev 0: II.1 Identify Limiting Increase in Heat Conforms None. 15.1.1-15.1.4 ease in Feedwater Basic Objective Removal Events perature, Increase in water Flow, Increase in m Flow, and Inadvertent ning of the Turbine ss System or Inadvertent ration of the Decay Heat oval System S 15.1.1-15.1.4, Rev 0: II.2 Verify Fuel Damage and System Conforms None. 15.1.1-15.1.4 ease in Feedwater Basic Objective Pressure Criteria are Met for Limiting perature, Increase in Event. water Flow, Increase in Conformance with Regulatory Criteria m Flow, and Inadvertent ning of the Turbine ss System or Inadvertent ration of the Decay Heat oval System

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.1.1-15.1.4, Rev 0: II.1 System Pressure Conforms None. 15.1.1-15.1.4 ease in Feedwater Specific perature, Increase in Criterion water Flow, Increase in m Flow, and Inadvertent ning of the Turbine ss System or Inadvertent ration of the Decay Heat oval System S 15.1.1-15.1.4, Rev 0: II.2 MCHFR Remains Above 95/95 Limit Conforms NuScale has determined that critical heat 15.1.1-15.1.4 ease in Feedwater Specific flux more accurately describes plant perature, Increase in Criterion phenomena than departure from nucleate water Flow, Increase in boiling. m Flow, and Inadvertent ning of the Turbine ss System or Inadvertent ration of the Decay Heat oval System S 15.1.1-15.1.4, Rev 0: II.3 AOOs Should Not Generate More Conforms NuScale events are classified by AOO, IE, 15.1.1-15.1.4 ease in Feedwater Specific Serious Condition accident, and special event, but will perature, Increase in Criterion conform with SRP requirement that water Flow, Increase in incidents of moderate frequency should not m Flow, and Inadvertent generate a more serious plant condition ning of the Turbine without other faults occurring ss System or Inadvertent independently. Conformance with Regulatory Criteria ration of the Decay Heat oval System S 15.1.1-15.1.4, Rev 0: II.4 Instrument Spans and Setpoints use Conforms None. 15.1.1-15.1.4 ease in Feedwater Specific RG 1.105 perature, Increase in Criterion water Flow, Increase in m Flow, and Inadvertent ning of the Turbine ss System or Inadvertent ration of the Decay Heat oval System

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.1.1-15.1.4, Rev 0: II.5 Identify Limiting Single Failure Conforms None. 15.1.1-15.1.4 ease in Feedwater Specific perature, Increase in Criterion water Flow, Increase in m Flow, and Inadvertent ning of the Turbine ss System or Inadvertent ration of the Decay Heat oval System S 15.1.1-15.1.4, Rev 0: II.1 Analytical Initial Power Level is 102% Conforms None. 15.1.1-15.1.4 ease in Feedwater Parameters perature, Increase in water Flow, Increase in m Flow, and Inadvertent ning of the Turbine ss System or Inadvertent ration of the Decay Heat oval System S 15.1.1-15.1.4, Rev 0: II.2 Analytical Conservative Scram Used Conforms None. 15.1.1-15.1.4 ease in Feedwater Parameters perature, Increase in water Flow, Increase in m Flow, and Inadvertent ning of the Turbine ss System or Inadvertent Conformance with Regulatory Criteria ration of the Decay Heat oval System S 15.1.1-15.1.4, Rev 0: II.3 Analytical Core Burnup Conforms None. 15.1.1-15.1.4 ease in Feedwater Parameters perature, Increase in water Flow, Increase in m Flow, and Inadvertent ning of the Turbine ss System or Inadvertent ration of the Decay Heat oval System

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.1.1-15.1.4, Rev 0: II.4 Analytical Setpoint Inaccuracies use guidance in Conforms None. 15.1.1-15.1.4 ease in Feedwater Parameters RG 1.105 perature, Increase in water Flow, Increase in m Flow, and Inadvertent ning of the Turbine ss System or Inadvertent ration of the Decay Heat oval System S 15.1.5, Rev 0: Steam II.1 Reactor Coolant and Main Steam Conforms None. 15.1.5 em Piping Failures Inside Specific System Pressure Outside of Containment Criteria S 15.1.5, Rev 0: Steam II.2 Evaluation of Core Damage Potential Conforms NuScale has determined that critical heat 15.1.5 em Piping Failures Inside Specific flux more accurately describes plant Outside of Containment Criteria phenomena than departure from nucleate boiling. S 15.1.5, Rev 0: Steam II.3 Radiological Criteria for Steam Line Conforms None. 15.1.5 em Piping Failures Inside Specific Breaks Outside of Containment Criteria S 15.1.5, Rev 0: Steam II.4 Safety-Related Classification and Conforms None. 15.1.5 em Piping Failures Inside Specific Auto-Initiation of Decay Heat Outside of Containment Criteria Removal System S 15.1.5, Rev 0: Steam II.1 Initial Power Level and Plant Conforms None. 15.1.5 em Piping Failures Inside Assumptions Operating Mode Outside of Containment Conformance with Regulatory Criteria S 15.1.5, Rev 0: Steam II.2 Loss of Offsite Power Conforms None. 15.1.5 em Piping Failures Inside Assumptions Outside of Containment S 15.1.5, Rev 0: Steam II.3 Postulated Steam Line Break Effects Conforms None. 15.1.5 em Piping Failures Inside Assumptions Outside of Containment S 15.1.5, Rev 0: Steam II.4 Worst Case Failure of Single Active Conforms None. 15.1.5 em Piping Failures Inside Assumptions Component Outside of Containment

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.1.5, Rev 0: Steam II.5 Maximum-Worth Rod Fully Conforms None. 15.1.5 em Piping Failures Inside Assumptions Withdrawn Outside of Containment S 15.1.5, Rev 0: Steam II.6 Core Burnup Conforms None. 15.1.5 em Piping Failures Inside Assumptions Outside of Containment S 15.1.5, Rev 0: Steam II.7 Initial Core Flow Conforms NuScale has determined that critical heat 15.1.5 em Piping Failures Inside Assumptions flux more accurately describes plant Outside of Containment phenomena than departure from nucleate boiling. S 15.1.5, Rev 0: Steam II.8 Postulated Failure of Non-Seismic Conforms None. 15.1.5 em Piping Failures Inside Assumptions Main Steam Line Outside of Containment S 15.1.5, Rev 0: Steam II.9 Postulated Failure of Seismic Main Conforms None. 15.1.5 em Piping Failures Inside Assumptions Steam Line Outside of Containment S 15.1.5, Rev 0: Steam II.10 Limiting Consequence Assessment Not Applicable Operator Action is not required to mitigate Not Applicable em Piping Failures Inside Assumptions When Operator Action is Credited the consequences of a steam line break. Outside of Containment 15.1.5.A, Rev 2: All Various Partially Conforms Per SRP Section 15.0.3, Section I, Areas of 15.0.3 ological Consequences Review, Item 10 under subheading Review ain Steam Line Failures Interfaces, for the review of design ide Containment of a certification applications, SRP Section 15.0.3 supersedes the radiological analyses, assumptions, acceptance criteria, and Conformance with Regulatory Criteria methodologies identified in this SRP Section 15.1.5, Appendix A. Provisions related to the nonradiological analyses aspects of this SRP Section 15.1.5, Appendix A, remain applicable to the DCA.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.1.5.A, Rev 2: II (No number) First full paragraph and Items 1 and 2 Partially Conforms The part of this guidance specifying the 15.0.3 ological Consequences on Page 15.1.5-11, Exposure calculation of radiological consequences of ain Steam Line Failures Guidelines for Calculated Doses a postulated main steam line break outside ide Containment of a containment is applicable to the DCA. However, per SRP Section 15.0.3, Section I, Areas of Review, Item 10.A under subheading Review Interfaces, the part of this acceptance criterion that specifies radiological acceptance criteria is superseded by SRP Section 15.0.3. 15.1.5.A, Rev 2: II (No number) First full paragraph following Items 1 Not Applicable This acceptance criterion specifies Not Applicable ological Consequences and 2 on Page 15.1.5-11, radiological analysis methodology and ain Steam Line Failures Methodology and Assumptions for assumptions that are superseded by SRP ide Containment of a Calculating Radiological Section 15.0.3. Consequences 15.1.5.A, Rev 2: II (No number) Second full paragraph following Partially Conforms The part of this guidance related to the 15.0.3 ological Consequences Items 1 and 2 on Page 15.1.5-11, required technical specification for primary ain Steam Line Failures Technical Specifications for Assumed and secondary coolant iodine activity and ide Containment of a Iodine Activity and Primary-to- primary-to-secondary leak rate is applicable Secondary Leak Rate to the DCA. However, per SRP Section 15.0.3, Section I, Areas of Review, Item 10.A under subheading Review Interfaces, the part of this acceptance criterion that specifies radiological acceptance criteria is superseded by SRP Section 15.0.3. Conformance with Regulatory Criteria S 15.1.6, Rev 0: Loss of II.1 Specific Reactor Coolant Pressure Conforms None. 15.1.6 ainment Vacuum Criteria S 15.1.6, Rev 0: Loss of II.2 Specific Cladding Integrity Conforms None. 15.1.6 ainment Vacuum Criteria S 15.1.6, Rev 0: Loss of II.3 Specific AOOs Should Not Generate More Conforms None. 15.1.6 ainment Vacuum Criteria Serious Condition S 15.1.6, Rev 0: Loss of II.4 Specific Instruments Spans and Setpoints use Conforms None. 15.1.6 ainment Vacuum Criteria RG 1.105 S 15.1.6, Rev 0: Loss of II.5 Specific Identify Limiting Single Failure Conforms None. 15.1.6 ainment Vacuum Criteria

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.1.6, Rev 0: Loss of II.1 Analytical Initial Power Level is 102% Conforms None. 15.1.6 ainment Vacuum Parameters S 15.1.6, Rev 0: Loss of II.2 Analytical Conservative Scram Used Conforms None. 15.1.6 ainment Vacuum Parameters S 15.1.6, Rev 0: Loss of II.3 Analytical Core Burnup Conforms None. 15.1.6 ainment Vacuum Parameters S 15.1.6, Rev 0: Loss of II.4 Analytical Maximize Heat Transfer from RCS to Conforms None. 15.1.6 ainment Vacuum Parameters Containment and Reactor Pool S 15.1.6, Rev 0: Loss of II.5 Analytical Setpoint Inaccuracies use Guidance in Conforms None. 15.1.6 ainment Vacuum Parameters RG 1.105 S 15.2.1-15.2.5, Rev 0: II.1 Basic Objectives - Initiating Events Conforms None. 15.2.1-15.2.5 of External Load; Turbine Loss of Condenser um; Closure of Main m Isolation Valve; and m Pressure Regulator re (Closed) S 15.2.1-15.2.5, Rev 0: II.2 Specific Criteria for Events of Partially Conforms The NuScale design does not have a steam 15.2.1-15.2.5 of External Load; Turbine Moderate Frequency pressure regulator. Loss of Condenser um; Closure of Main m Isolation Valve; and m Pressure Regulator re (Closed) Conformance with Regulatory Criteria S 15.2.1-15.2.5, Rev 0: II.2.A Reactor Coolant System and Main Conforms None. 15.2.1-15.2.5 of External Load; Turbine Steam System Pressures Loss of Condenser um; Closure of Main m Isolation Valve; and m Pressure Regulator re (Closed)

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.2.1-15.2.5, Rev 0: II.2.B Fuel Cladding Integrity Conforms The NuScale design does not have a steam 15.2.1-15.2.5 of External Load; Turbine pressure regulator. Loss of Condenser um; Closure of Main m Isolation Valve; and m Pressure Regulator re (Closed) S 15.2.1-15.2.5, Rev 0: II.2.C Incidents of Moderate Frequency Conforms None. 15.2.1-15.2.5 of External Load; Turbine Loss of Condenser um; Closure of Main m Isolation Valve; and m Pressure Regulator re (Closed) S 15.2.1-15.2.5, Rev 0: II.2.D Instrument Setpoints - Impact on Conforms None. 15.2.1-15.2.5 of External Load; Turbine Plant Response Loss of Condenser um; Closure of Main m Isolation Valve; and m Pressure Regulator re (Closed) S 15.2.1-15.2.5, Rev 0: II.2.E Most Limiting Plant System Single Conforms None. 15.2.1-15.2.5 of External Load; Turbine Failure Loss of Condenser um; Closure of Main Conformance with Regulatory Criteria m Isolation Valve; and m Pressure Regulator re (Closed) S 15.2.1-15.2.5, Rev 0: II.2.F Performance of Nonsafety-Related Conforms None. 15.2.1-15.2.5 of External Load; Turbine Systems and Single Failures of Active Loss of Condenser and Passive Systems um; Closure of Main m Isolation Valve; and m Pressure Regulator re (Closed)

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.2.1-15.2.5, Rev 0: II.3 Analytical Model Conforms None. 15.2.1-15.2.5 of External Load; Turbine Loss of Condenser um; Closure of Main m Isolation Valve; and m Pressure Regulator re (Closed) S 15.2.1-15.2.5, Rev 0: II.3.A Values of Parameters Used in Conforms None. 15.2.1-15.2.5 of External Load; Turbine Analytical Model - Initial Power Level Loss of Condenser and Modes of Operation um; Closure of Main m Isolation Valve; and m Pressure Regulator re (Closed) S 15.2.1-15.2.5, Rev 0: II.3.B Values of Parameters Used in Conforms None. 15.2.1-15.2.5 of External Load; Turbine Analytical Model - Scram Loss of Condenser Characteristics um; Closure of Main m Isolation Valve; and m Pressure Regulator re (Closed) S 15.2.1-15.2.5, Rev 0: II.3.C Values of Parameters Used in Conforms None. 15.2.1-15.2.5 of External Load; Turbine Analytical Model - Core Burnup Loss of Condenser um; Closure of Main Conformance with Regulatory Criteria m Isolation Valve; and m Pressure Regulator re (Closed) S 15.2.1-15.2.5, Rev 0: II.3.D Values of Parameters Used in Conforms None. 15.2.1-15.2.5 of External Load; Turbine Analytical Model - Instrumentation Loss of Condenser Setpoints for Mitigating System um; Closure of Main Actuation m Isolation Valve; and m Pressure Regulator re (Closed)

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.2.6, Rev 0: Loss of II.1 Reactor Coolant and Main Steam Conforms None. 15.2.6 emergency AC Power to System Pressures tation Auxiliaries S 15.2.6, Rev 0: Loss of II.2 Fuel Cladding Integrity Conforms None. 15.2.6 emergency AC Power to tation Auxiliaries S 15.2.6, Rev 0: Loss of II.3 Incidents of Moderate Frequency Conforms NuScale categorizes events as AOO and IE. 15.2.6 emergency AC Power to tation Auxiliaries S 15.2.6, Rev 0: Loss of II.4 Requirements of GDC 10 and GDC 15 Conforms None. 15.2.6 emergency AC Power to tation Auxiliaries S 15.2.6, Rev 0: Loss of II.5 Most Limiting Plant System Single Conforms None. 15.2.6 emergency AC Power to Failure tation Auxiliaries S 15.2.6, Rev 0: Loss of II.5 A-D Analysis of Loss of AC Power - Conforms None. 15.2.6 emergency AC Power to Analytical Model and Methods, tation Auxiliaries conservative assumptions and RG 1.105 S 15.2.7, Rev 0: Loss of II.1 Fuel and System Pressure Parameters Conforms None. 15.2.7 mal Feedwater Flow met S 15.2.7, Rev 0: Loss of II.2 Events of Moderate Frequency Conforms NuScale categorizes events as AOO and IE. 15.2.7 mal Feedwater Flow S 15.2.7, Rev 0: Loss of II.3 Analytical Model and Methods Conforms None. 15.2.7 Conformance with Regulatory Criteria mal Feedwater Flow S 15.2.8, Rev 0: Feedwater II.1 Reactor Coolant System and Main Conforms None. 15.2.8 em Pipe Break Inside and Steam System Pressures ide Containment (PWR) S 15.2.8, Rev 0: Feedwater II.2 Evaluation of Core Damage Potential Conforms For slower reactivity insertions, NuScale 15.2.8 em Pipe Break Inside and uses a heat generation rate limit to ensure ide Containment (PWR) that fuel centerline melting limits are met. S 15.2.8, Rev 0: Feedwater II.3 Calculated Site Boundary Doses Conforms None. 15.2.8 em Pipe Break Inside and ide Containment (PWR)

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.2.8, Rev 0: Feedwater II.4 DHRS must be safety grade and Conforms None. 15.2.8 em Pipe Break Inside and automatically initiated when ide Containment (PWR) required. S 15.2.8, Rev 0: Feedwater II.5 Assumptions for Initial Plant Conforms None. 15.2.8 em Pipe Break Inside and Conditions and Postulated Failures ide Containment (PWR) 15.3.1-15.3.2, Rev 2: Loss All Various Not Applicable Applicable only to LWR designs that rely on Not Applicable rced Reactor Coolant forced reactor coolant flow for core cooling. Including Trip of Pump The NuScale design uses passive natural or and Flow Controller circulation of the primary coolant, unctions eliminating the need for reactor coolant pumps. 15.3.3-15.3.4, Rev 2: All Various Not Applicable Section 15.3.3 - 15.3.4 are applicable only to Not Applicable tor Coolant Pump Rotor LWR designs that rely on forced reactor ure and Reactor Coolant coolant flow for core cooling. The NuScale p Shaft Break design uses passive natural circulation of the primary coolant, eliminating the need for reactor coolant pumps. 15.4.1, Rev 3: II.1.A Thermal Margin Limits Conforms Critical heat flux (CHF) is more appropriate 15.4.1 ontrolled Control Rod terminology for NuScale phenomena than mbly Withdrawal From a departure from nucleate boiling (DNBR). ritical or Low Power up Condition 15.4.1, Rev 3: II.1.B Fuel Centerline Temperatures Conforms For slower reactivity insertions, NuScale 15.4.1 ontrolled Control Rod uses a heat generation rate limit to ensure Conformance with Regulatory Criteria mbly Withdrawal From a that fuel centerline melting limits are met. ritical or Low Power up Condition 15.4.1, Rev 3: II.1.C Uniform Cladding Strain Not Applicable The SRP states that this criterion applies to Not Applicable ontrolled Control Rod BWRs. NuScale uses the 95/95 MCHFR mbly Withdrawal From a approach to ensure no cladding or fuel ritical or Low Power failures. up Condition 15.4.2, Rev 3: II.1.A Thermal Margin Limits Conforms CHF is more appropriate terminology for 15.4.2 ontrolled Control Rod NuScale phenomenon than DNBR. mbly Withdrawal at er

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.4.2, Rev 3: II.1.B Fuel Centerline Temperatures Conforms For slower reactivity insertions, NuScale 15.4.2 ontrolled Control Rod uses a heat generation rate limit to ensure mbly Withdrawal at that fuel centerline melting limits are met. er 15.4.2, Rev 3: II.1.C Uniform Cladding Strain Not Applicable The SRP states that this criterion applies to Not Applicable ontrolled Control Rod BWRs. NuScale uses the 95/95 MCHFR mbly Withdrawal at approach to ensure no cladding or fuel er failures. 15.4.3, Rev 3: Control Rod II.1 Thermal Margin Limits Conforms CHF is more appropriate terminology for 15.4.3 peration (System NuScale phenomenon than DNBR. unction or Operator

 )

15.4.3, Rev 3: Control Rod II.2 Fuel Centerline Temperatures Conforms For slower reactivity insertions, NuScale 15.4.3 peration (System uses a heat generation rate limit to meet unction or Operator fuel centerline melting limits.

 )

15.4.3, Rev 3: Control Rod II.3 Uniform Cladding Strain Conforms None. 15.4.3 peration (System unction or Operator

 )

15.4.4-15.4.5, Rev 2: II.A RCS and MSS Pressures Not Applicable This guidance is not applicable because the Not Applicable up of an Inactive Loop or specific language refers to PWR designs that culation Loop at an use forced reactor coolant flow and have rrect Temperature, and reactor coolant loops and pumps. The Controller Malfunction NuScale design does not require or include Conformance with Regulatory Criteria ing an Increase in BWR reactor coolant pumps. The potential for a Flow Rate postulated startup reactivity accident (e.g., initiated by abnormal startup sequence) has been identified as an event requiring consideration for the NuScale reactor design. The guidance for these AOOs is available in DSRS 15.4.6 - Inadvertent Decrease in Boron Concentration in the Reactor Coolant System (PWR).

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.4.4-15.4.5, Rev 2: II.B Fuel Thermal Limits Not Applicable This guidance is not applicable because the Not Applicable up of an Inactive Loop or specific language refers to PWR designs that culation Loop at an use forced reactor coolant flow and have rrect Temperature, and reactor coolant loops and pumps. The Controller Malfunction NuScale design does not require or include ing an Increase in BWR reactor coolant pumps. The potential for a Flow Rate postulated startup reactivity accident (e.g., initiated by abnormal startup sequence) has been identified as an event requiring consideration for the NuScale reactor design. The guidance for these AOOs is available in DSRS 15.4.6 - Inadvertent Decrease in Boron Concentration in the Reactor Coolant System (PWR). 15.4.4-15.4.5, Rev 2: II.C Events of Moderate Frequency Not Applicable This guidance is not applicable because the Not Applicable up of an Inactive Loop or specific language refers to PWR designs that culation Loop at an use forced reactor coolant flow and have rrect Temperature, and reactor coolant loops and pumps. The Controller Malfunction NuScale design does not require or include ing an Increase in BWR reactor coolant pumps. The potential for a Flow Rate postulated startup reactivity accident (e.g., initiated by abnormal startup sequence) has been identified as an event requiring consideration for the NuScale reactor design. The guidance for these AOOs is Conformance with Regulatory Criteria available in DSRS 15.4.6 - Inadvertent Decrease in Boron Concentration in the Reactor Coolant System (PWR).

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.4.4-15.4.5, Rev 2: II.D Instrument Setpoints Not Applicable This guidance is not applicable because the Not Applicable up of an Inactive Loop or specific language refers to PWR designs that culation Loop at an use forced reactor coolant flow and have rrect Temperature, and reactor coolant loops and pumps. The Controller Malfunction NuScale design does not require or include ing an Increase in BWR reactor coolant pumps. The potential for a Flow Rate postulated startup reactivity accident (e.g., initiated by abnormal startup sequence) has been identified as an event requiring consideration for the NuScale reactor design. The guidance for these AOOs is available in DSRS 15.4.6 - Inadvertent Decrease in Boron Concentration in the Reactor Coolant System (PWR). 15.4.4-15.4.5, Rev 2: II.E Single Failure Not Applicable This guidance is not applicable because the Not Applicable up of an Inactive Loop or specific language refers to PWR designs that culation Loop at an use forced reactor coolant flow and have rrect Temperature, and reactor coolant loops and pumps. The Controller Malfunction NuScale design does not require or include ing an Increase in BWR reactor coolant pumps. The potential for a Flow Rate postulated startup reactivity accident (e.g., initiated by abnormal startup sequence) has been identified as an event requiring consideration for the NuScale reactor design. The guidance for these AOOs is Conformance with Regulatory Criteria available in DSRS 15.4.6 - Inadvertent Decrease in Boron Concentration in the Reactor Coolant System (PWR).

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.4.4-15.4.5, Rev 2: II.F Non-Safety Systems Not Applicable This guidance is not applicable because the Not Applicable up of an Inactive Loop or specific language refers to PWR designs that culation Loop at an use forced reactor coolant flow and have rrect Temperature, and reactor coolant loops and pumps. The Controller Malfunction NuScale design does not require or include ing an Increase in BWR reactor coolant pumps. The potential for a Flow Rate postulated startup reactivity accident (e.g., initiated by abnormal startup sequence) has been identified as an event requiring consideration for the NuScale reactor design. The guidance for these AOOs is available in DSRS 15.4.6 - Inadvertent Decrease in Boron Concentration in the Reactor Coolant System (PWR). 15.4.6, Rev 2: Inadvertent II.1 Reactor Coolant and Main Steam Conforms None. 15.4.6 ease in Boron System Pressures centration in the Reactor ant System (PWR) 15.4.6, Rev 2: Inadvertent II.2 Fuel Cladding Integrity Conforms CHF is more appropriate terminology for 15.4.6 ease in Boron NuScale phenomenon. centration in the Reactor ant System (PWR) 15.4.6, Rev 2: Inadvertent II.3 Incidents of Moderate Frequency Conforms None. 15.4.6 ease in Boron centration in the Reactor Conformance with Regulatory Criteria ant System (PWR) 15.4.6, Rev 2: Inadvertent II.4 Minimum Time Intervals for Required Not Applicable Operator action is not required to mitigate Not Applicable ease in Boron Operator Actions an inadvertent boron dilution event. centration in the Reactor ant System (PWR) 15.4.6, Rev 2: Inadvertent II.5 Analysis Model, Methods, and Conforms None. 15.4.6 ease in Boron Assumptions centration in the Reactor ant System (PWR)

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.4.7, Rev 2: Inadvertent II.1 Provision in Plant Operating Conforms None. 15.4.7 ing and Operation of a Procedures Requiring Assembly in an Improper Instrumentation to Detect Fuel ion Loading Errors 15.4.7, Rev 2: Inadvertent II.2 Offsite Radiological Consequences Conforms Safety analysis demonstrates that there are 15.4.7 ing and Operation of a no fuel failures. Assembly in an Improper ion 15.4.8, Rev 3: Spectrum of II.1 Availability of Monitoring Conforms None. 15.4.8 Ejection Accidents (PWR) Instrumentation 15.4.8, Rev 3: Spectrum of II.2 Effects of Postulated Reactivity Conforms None. 15.4.8 Ejection Accidents (PWR) Accidents 15.4.8, Rev 3: Spectrum of II.3 Radiation Dose Limits Conforms None. 15.4.8 Ejection Accidents (PWR) 15.4.8.A, Rev 1: All Various Partially Conforms Per SRP Section 15.0.3, Section I, Areas of 15.0.3 ological Consequences Review, Item 10 under subheading Review Control Rod Ejection Interfaces, for the review of design dent (PWR) certification applications, SRP Section 15.0.3 supersedes the radiological analyses, assumptions, acceptance criteria, and methodologies identified in SRP Section 15.4.8, Appendix A. Provisions related to the nonradiological analyses aspects of this SRP Section 15.4.8, Appendix A, apply to the DCA. Conformance with Regulatory Criteria 15.4.8.A, Rev 1: II (No number) First paragraph of Section II (bottom Partially Conforms The part of this guidance specifying the 15.0.3 ological Consequences of page 15.4.8-5 and top of page calculation of radiological consequences of Control Rod Ejection 15.4.8-6) - Acceptability of Site and a postulated control rod ejection accident is dent (PWR) Dose Mitigating ESF applicable to the DCA. However, per SRP Section 15.0.3, Section I, Areas of Review, Item 10.C under subheading Review Interfaces, the part of this acceptance criterion that specifies radiological acceptance criteria is superseded by SRP Section 15.0.3.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.4.8.A, Rev 1: II (No number) First full paragraph on page 15.4.8-6) Partially Conforms The part of this guidance related to the 15.0.3 ological Consequences - Technical Specification for Primary- required technical specification for primary-Control Rod Ejection to-Secondary Leak Rate to-secondary leak rate is applicable to the dent (PWR) DCA. However, per SRP Section 15.0.3, Section I, Areas of Review, Item 10.C under subheading Review Interfaces, the part of this acceptance criterion that specifies radiological acceptance criteria is superseded by SRP Section 15.0.3. 15.4.8.A, Rev 1: II (No number) Second full paragraph on page Not Applicable Per SRP Section 15.0.3, Section I, Areas of Not Applicable ological Consequences 15.4.8-6) - Dose Model Review, Item 10.C under subheading Control Rod Ejection Review Interfaces, this acceptance criterion dent (PWR) specifies radiological acceptance criteria and assumptions that are superseded by SRP Section 15.0.3. 15.4.9, Rev 3: Spectrum of All - Not Applicable This SRP section and its acceptance criteria Not Applicable Drop Accidents (BWR) (II.1 through II.3) are applicable only to BWRs. 15.4.9.A, Draft Rev 3: All - Not Applicable This SRP section and its acceptance criteria Not Applicable ological Consequences are applicable only to BWRs. ntrol Rod Drop Accident R) S 15.5.1-15.5.2, Rev 0: II.1 The frequency classification for this Conforms None. 15.0 mical and Volume event is an AOO. 15.5.1-15.5.2 rol System Malfunction Conformance with Regulatory Criteria Increases Reactor ant Inventory S 15.5.1-15.5.2, Rev 0: II.2 The sequence of events, from Conforms None. 15.5.1-15.5.2 mical and Volume initiation until a stabilized condition rol System Malfunction is reached including assumptions for Increases Reactor equipment that operates, fails to ant Inventory operate or requires operator action. S 15.5.1-15.5.2, Rev 0: II.3 Evaluation Model must be an Conforms None. 15.5.1-15.5.2 mical and Volume approved model or be justified. rol System Malfunction Increases Reactor ant Inventory

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.5.1-15.5.2, Rev 0: II.4.A Input Parameters and Initial Conforms None. 15.5.1-15.5.2 mical and Volume Conditions - Initial Power Level rol System Malfunction Increases Reactor ant Inventory S 15.5.1-15.5.2, Rev 0: II.4.B Input Parameters and Initial Conforms None. 15.5.1-15.5.2 mical and Volume Conditions - Scram Characteristics rol System Malfunction Increases Reactor ant Inventory S 15.5.1-15.5.2, Rev 0: II.4.C Input Parameters and Initial Conforms None. 15.5.1-15.5.2 mical and Volume Conditions - Core Burnup rol System Malfunction Increases Reactor ant Inventory 15.6.1, Rev 2: Inadvertent All Various Partially Conforms This guidance is only applicable to LWRs 15.6.1 ning of a PWR Pressurizer that are designed with power-operated 15.6.6 f Valve or a BWR Pressure pressurizer relief valves. The NuScale design f Valve does not use power-operated relief valves (PORVs), which have the potential to open inadvertently. Rather, the NuScale design uses springloaded ASME code safety relief valves, which do not have the PORVs vulnerability to inadvertent operation. However, a mechanical failure of the reactor Conformance with Regulatory Criteria safety valve (RSV) is bounded by an inadvertent ECCS valve actuation, analyzed in Section 15.6.6.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.6.2, Rev 2: Radiological II (No Number) Penultimate paragraph of Section II Partially Conforms The part of this guidance specifying the 15.0.3 equences of the Failure on page 15.6.22 - Acceptability of Site calculation of radiological consequences of 15.6.2 mall Lines Carrying and Dose Mitigating ESF Systems a postulated failure outside containment of ary Coolant Outside a small reactor coolant line is applicable to ainment the DCA. However, per SRP Section 15.0.3, Section I, Areas of Review, Item 10.E under subheading Review Interfaces, the part of this acceptance criterion that specifies radiological acceptance criteria is superseded by SRP Section 15.0.3. 15.6.2, Rev 2: Radiological II (No Number) Last paragraph of Section II on page Partially Conforms The part of this guidance related to the 15.0.3 equences of the Failure 15.6.22 - Plant-Specific Technical required technical specification for primary 15.6.2 mall Lines Carrying Specifications for Primary Coolant coolant iodine activity is applicable to the ary Coolant Outside System Iodine Activity DCA. However, per SRP Section 15.0.3, ainment Section I, Areas of Review, Item 10.E under subheading Review Interfaces, the part of this acceptance criterion that specifies radiological acceptance criteria is superseded by SRP Section 15.0.3. 15.6.3, Rev 2: Radiological II (No Number) First paragraph and Items (1) and (2) Partially Conforms The part of this guidance specifying the 15.0.3 equences of Steam of Section II on page 15.6.32 - calculation of radiological consequences of 15.6.3 erator Tube Failure (PWR) Acceptability of Site and Dose a postulated steam generator tube failure is Mitigating ESF Systems applicable to the DCA. However, per SRP Section 15.0.3, Section I, Areas of Review, Item 10.F under subheading Review Conformance with Regulatory Criteria Interfaces, the part of this acceptance criterion that specifies radiological acceptance criteria is superseded by SRP Section 15.0.3. 15.6.3, Rev 2: Radiological II (No Number) First sentence of the last paragraph of Not Applicable This acceptance criterion specifies Not Applicable equences of Steam Section II on page 15.6.32 - radiological analysis methodology and erator Tube Failure (PWR) Methodology and Assumptions for assumptions that are superseded by SRP Calculating Radiological Section 15.0.3. Consequences

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.6.3, Rev 2: Radiological II (No Number) Last two sentences of the last Partially Conforms The part of this guidance related to the 15.0.3 equences of Steam paragraph of Section II on page required technical specification for primary 15.6.3 erator Tube Failure (PWR) 15.6.32 - Plant-Specific Technical and secondary coolant iodine activity is Specifications for Primary and applicable to the DCA. However, per SRP Secondary Coolant System Iodine Section 15.0.3, Section I, Areas of Review, Activity Item 10.F under subheading Review Interfaces, the part of this acceptance criterion that specifies radiological acceptance criteria is superseded by SRP Section 15.0.3. 15.6.4, Rev 2: Radiological All - Not Applicable This SRP section and its acceptance criteria Not Applicable equences of Main Steam are applicable only to BWRs. Failure Outside ainment (BWR) S 15.6.5, Rev 0: Loss-of- II.1 Evaluation of ECCS Performance Departure The NuScale design supports an exemption 15.6.5 ant Accidents Resulting from selected portions of 10 CFR 50 Spectrum of Postulated Appendix K. The features of Appendix K g Breaks within the requirements that are technically relevant tor Coolant Pressure to the NuScale design are included in the ndary Appendix K analysis of small break LOCAs. S 15.6.5, Rev 0: Loss-of- II.2 Radiological Consequences of Most Conforms Per SRP Section 15.0.3, Section I, Areas of 15.6.5 ant Accidents Resulting Severe LOCA Review, Item 10 under subheading Review Spectrum of Postulated Interfaces, for the review of design g Breaks within the certification applications, SRP Section 15.0.3 tor Coolant Pressure supersedes the radiological analyses, Conformance with Regulatory Criteria ndary assumptions, acceptance criteria, and methodologies identified in this SRP Section 15.6.5. S 15.6.5, Rev 0: Loss-of- II.3 TMI Action Plan Requirements Conforms None. 15.6.5 ant Accidents Resulting Spectrum of Postulated g Breaks within the tor Coolant Pressure ndary

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.6.5, Rev 0: Loss-of- II.4 Programmatic Requirements Conforms None. 15.6.5 ant Accidents Resulting Spectrum of Postulated g Breaks within the tor Coolant Pressure ndary 15.6.5.A, Rev 1: II.1 Calculated Doses and Containment Partially Conforms The part of this guidance specifying the 15.0.3 ological Consequences Leakage Contribution calculation of radiological consequences of 15.6.5 Design Basis Loss-of- a hypothetical LOCA is applicable to the ant Accident Including DCA. However, per SRP Section 15.0.3, ainment Leakage Section I, Areas of Review, Item 10.H under ribution subheading Review Interfaces, the part of this acceptance criterion that specifies radiological acceptance criteria is superseded by SRP Section 15.0.3. A portion of this guidance is applicable only to BWRs. 15.6.5.A, Rev 1: II.2 Model for and Calculation of Post- Partially Conforms The part of this guidance specifying the 15.0.3 ological Consequences LOCA Containment Leakage calculation of post LOCA containment 15.6.5 Design Basis Loss-of- Contribution leakage contribution is applicable to the ant Accident Including DCA. However, per SRP Section 15.0.3, ainment Leakage Section I, Areas of Review, Item 10.H under ribution subheading Review Interfaces, the part of this Acceptance Criterion that specifies radiological acceptance criteria and analysis model is superseded by SRP Section 15.0.3. Conformance with Regulatory Criteria A portion of this guidance is applicable only to BWRs. 15.6.5.B, Rev 1: II.1 ESF System Leakage Assumptions Conforms None. 15.0.3 ological Consequences 15.6.5 Design Basis Loss-of-ant Accident: Leakage Engineered Safety ure Components Outside ainment

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.6.5.B, Rev 1: II.2 Calculation of Radiological Partially Conforms The part of this guidance specifying the 15.0.3 ological Consequences Consequences calculation of radiological consequences of 15.6.5 Design Basis Loss-of- postulated leakage is applicable to the DCA. ant Accident: Leakage However, per SRP Section 15.0.3, Section I, Engineered Safety Areas of Review, Item 10.H under ure Components Outside subheading Review Interfaces, the part of ainment this acceptance criterion that specifies radiological analyses, assumptions, acceptance criteria, and methodologies is superseded by SRP Section 15.0.3. 15.6.5.B, Rev 1: II.3 Combining Radiological Partially Conforms The part of this guidance specifying that 15.0.3 ological Consequences Consequences radiological consequences from ESF 15.6.5 Design Basis Loss-of- component leakage should be combined ant Accident: Leakage with consequences from other fission Engineered Safety product release paths is applicable to the ure Components Outside DCA. However, per SRP Section 15.0.3, ainment Section I, Areas of Review, Item 10.H under subheading Review Interfaces, the part of this acceptance criterion that specifies radiological acceptance criteria is superseded by SRP Section 15.0.3. A portion of this guidance is applicable only to BWRs. 15.6.5.D, Rev 1: All Various Not Applicable This SRP section and its acceptance criteria Not Applicable ological Consequences are applicable only to BWRs. Design Basis Loss-of-Conformance with Regulatory Criteria ant Accident: Leakage Main Steam Isolation e Leakage Control em (BWR) S 15.6.6, Rev 0: II.1 RCS pressure below 110 percent Conforms None. 15.6.6 vertent Operation of design value. S S 15.6.6, Rev 0: II.2 Maintain minimum DNBR. Conforms NuScale evaluated CHF as it is more 15.6.6 vertent Operation of appropriate than DNBR for the NuScale S design.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.6.6, Rev 0: II.3 An AOO should not develop more Conforms None. 15.6.6 vertent Operation of serious plant condition without other S faults occurring independently. 15.7.3, Rev 2: Radioactive All Various Partially Conforms The technical content has been relocated to 11.2 ase from a Subsystem or Branch Technical Position 11-6, which is ponent referenced in Section 11.2. 15.7.4, Rev 1: Radiological II.1 Acceptability of Site and Dose Not Applicable This acceptance criterion specifies Not Applicable equences of Fuel Mitigating ESF Systems radiological analysis acceptance criteria dling Accidents that are superseded by SRP Section 15.0.3. 15.7.4, Rev 1: Radiological II.2 Radioactivity Control Features of Fuel Partially Conforms The portion of this acceptance criterion 15.7.4 equences of Fuel Storage and Handling Systems related to fuel storage and handling dling Accidents systems inside the Fuel Building is applicable to those systems inside the NuScale Reactor Building. The portion of this acceptance criterion related to fuel storage and handling systems inside containment is applicable only to large LWR designs that incorporate a containment building housing numerous plant SSC. The NuScale design does not use a containment building. Rather, each NPM has its own compact steel containment vessel. This containment vessel does not contain fuel storage and handling systems. Thus, the portion of this acceptance criterion related Conformance with Regulatory Criteria to fuel storage and handling systems inside containment is not applicable. 15.7.4, Rev 1: Radiological II.3 Dose Model and Modeling Not Applicable This acceptance criterion specifies Not Applicable equences of Fuel Assumptions radiological analysis methodology and dling Accidents assumptions that are superseded by SRP Section 15.0.3.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.7.4, Rev 1: Radiological II.4 ESF Grade Atmosphere Clean-Up Not Applicable The NuScale design does not rely on ESF Not Applicable equences of Fuel System in Spent Fuel Storage Area ventilation systems to mitigate the dling Accidents consequences of a design basis accident. Nonsafety-related normal ventilation systems provide atmosphere cleanup capability, as necessary, that meets the design, testing, and maintenance guidelines in RG 1.140. These systems provide appropriate containment, confinement, and filtering to limit releases of airborne radioactivity to the environment during normal operations, anticipated operational occurrences, and postulated accident conditions. However, these systems are not required following an accident, and receive no credit in the determination of the radiological consequences of an accident. 15.7.4, Rev 1: Radiological II.5 Radiation Detection in Containment Partially Conforms The intent of this acceptance criterion is 15.7.4 equences of Fuel applicable but the specific language refers dling Accidents to LWR designs that incorporate a containment building within which fuel handling operations are performed. The NuScale design does not use a containment building. Rather, each NPM has its own Conformance with Regulatory Criteria compact steel containment vessel immediately surrounding the reactor vessel. The containment design provisions of this guidance for fuel handling operations inside containment are not relevant to the NuScale containment vessel design. However, the intent of this acceptance criterion is appropriate to apply to the NuScale Reactor Building, where the operating NPMs reside in the reactor pool and fuel handling operations are performed.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.7.5, Rev 2: Spent Fuel All Various Partially Conforms One of the principal functions of the 15.7.5 Drop Accidents NuScale reactor building crane (RBC) is to 15.7.6 move spent fuel casks in the Reactor Building refueling area. The RBC system design conforms to the single-failure-proof guidelines of NUREG-0612 so that any credible failure of a single component will not result in the loss of capability to stop and hold a critical load. The single-failure-proof crane precludes the need to perform load drop evaluations and as a result no accident analysis has been performed to assess radiological consequences of a spent fuel cask drop accident or a NPM drop accident. 15.7.5, Rev 2: Spent Fuel II.1 Acceptability of Site and Dose Not Applicable The RBC system design conforms to the Not Applicable Drop Accidents Mitigating ESF Systems single-failure-proof guidelines of NUREG-0612 so that any credible failure of a single component will not result in the loss of capability to stop and hold a critical load. The single-failure-proof crane precludes the need to perform load drop evaluations and as a result no accident analysis has been performed to assess radiological consequences of a spent fuel cask drop Conformance with Regulatory Criteria accident or a NPM drop accident. 15.7.5, Rev 2: Spent Fuel II.2 Radioactivity Control Features of Fuel Not Applicable The RBC system design conforms to the Not Applicable Drop Accidents Storage and Handling Systems single-failure-proof guidelines of NUREG-0612 so that any credible failure of a single component will not result in the loss of capability to stop and hold a critical load. The single-failure-proof crane precludes the need to perform load drop evaluations and as a result no accident analysis has been performed to assess radiological consequences of a spent fuel cask drop accident or a NPM drop accident.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.7.5, Rev 2: Spent Fuel II.3 Dose Model and Modeling Not Applicable The RBC system design conforms to the Not Applicable Drop Accidents Assumptions single-failure-proof guidelines of NUREG-0612 so that any credible failure of a single component will not result in the loss of capability to stop and hold a critical load. The single-failure-proof crane precludes the need to perform load drop evaluations and as a result no accident analysis has been performed to assess radiological consequences of a spent fuel cask drop accident or a NPM drop accident. 15.7.5, Rev 2: Spent Fuel II.4 ESF Grade Atmosphere Clean-Up Not Applicable The RBC system design conforms to the Not Applicable Drop Accidents System in Spent Fuel Storage Area single-failure-proof guidelines of NUREG-0612 so that any credible failure of a single component will not result in the loss of capability to stop and hold a critical load. The single-failure-proof crane precludes the need to perform load drop evaluations and as a result no accident analysis has been performed to assess radiological consequences of a spent fuel cask drop accident or a NPM drop accident. 15.7.5, Rev 2: Spent Fuel II.5 Plant Design Features Eliminating Partially Conforms The RBC system design conforms to the 15.7.5 Drop Accidents Need for Calculation single-failure-proof guidelines of NUREG- 15.7.6 0612 so that any credible failure of a single Conformance with Regulatory Criteria component will not result in the loss of capability to stop and hold a critical load. The single-failure-proof crane precludes the need to perform load drop evaluations and as a result no accident analysis has been performed to assess radiological consequences of a spent fuel cask drop accident or a NPM drop accident. 15.8, Rev 2: Anticipated II.1 Acceptance Criteria for Boiling Water Not Applicable This guidance is only applicable to BWRs. Not Applicable sients without Scram Reactors (BWRs) 15.8, Rev 2: Anticipated II.2 Acceptance Criteria for Pressurized Not Applicable NuScale is characterized as an evolutionary Not Applicable sients without Scram Water Reactors (PWRs) plant (See the acceptance criteria in II.3).

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 15.8, Rev 2: Anticipated II.3.A.i Provide a diverse scram system Partially Conforms The NuScale design relies on diversity 15.8 sients without Scram within the module protection system (MPS) to reduce the risk associated with ATWS events. Internal diversity within the MPS is a simpler approach and meets the intent of the diverse scram elements of the ATWS Rule. 15.8, Rev 2: Anticipated II.3.A.ii or, Demonstrate that the ATWS event Not Applicable As discussed in the comment above for Not Applicable sients without Scram consequences are acceptable Acceptance Criteria II.3.A.i, the NuScale design relies on diversity within the RPS to reduce the risk associated with ATWS events. 15.8, Rev 2: Anticipated II.3.B Required Equipment Does Not Apply Conforms As discussed above in the comment for 15.8 sients without Scram to Design Acceptance Criteria II.2, the design features required by 10 CFR 50.62(C)(1) either do not apply to the NuScale design or are not required to reduce the risk from ATWS events. Internal diversity within the MPS is a simpler approach to addressing the diverse scram elements of the ATWS Rule and acceptance criteria II.3.A.ii. and II.3.C(2) for evolutionary plants. 15.8, Rev 2: Anticipated II.3.C Analysis Demonstrating the Failure Partially Conforms NuScale conforms to the second criterion 15.8 sients without Scram Probability of Failing the ATWS option of reducing the probability of a Success Criteria is Sufficiently Small failure to scram. This is achieved with a Conformance with Regulatory Criteria diverse RPS instead of a diverse scram system as discussed above. S 15.9.A, Rev 0: Thermal II.1 No requirements - None. - aulic Stability Review onsibilities S 15.9.A, Rev 0: Thermal II.2 Meeting Requirements of GDC 12 Conforms None. 4.4.7 aulic Stability Review onsibilities S 15.9.A, Rev 0: Thermal II.3 Detect and suppress system criteria Not Applicable Reactor trip signals prevent violation of CHF Not Applicable aulic Stability Review for demonstrating acceptable limits before flow instabilities can develop. onsibilities consequences of stability

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 15.9.A, Rev 0: Thermal II.4 Detect and Suppress Method: Not Applicable Exclusion zone option is not used in the Not Applicable aulic Stability Review Exclusion zone and buffer region NuScale design. Reactor trip signals prevent onsibilities methodology violation of CHF limits before unstable flow oscillations can develop. Protective action occurs prior to development of oscillation. S 15.9.A, Rev 0: Thermal II.5 Detect and Suppress Method Trip of Partially Conforms Existing reactor trip signals provide an 4.4.4 aulic Stability Review reactor before SAFDL violation exclusion zone that prevents violation of 4.4.7 onsibilities SAFDL limits from other causes, which is already more limiting than the exclusion zone needed to preclude flow instabilities. S 15.9.A, Rev 0: Thermal II.6 Backup options if licensing solutions Not Applicable Detect and Suppress options are not Not Applicable aulic Stability Review declared inoperable employed. Existing technical specifications onsibilities for RTS provide controls on allowable unavailabilities of protective trips. Backup options are not required. S 15.9.A, Rev 0: Thermal II.7 Criteria to determine the Partially Conforms RTS system trips reactor prior to conditions 4.4.7 aulic Stability Review acceptability of the D&S System that could initiate flow instabilities. onsibilities compliance with the requirements of Stabilities are not detected and suppressed. GDC 20 S 15.9.A, Rev 0: Thermal II.8 Detect and Suppress system to Not Applicable RTS system trips reactor prior to conditions Not Applicable aulic Stability Review monitor process variables and that could initiate flow instabilities. onsibilities systems. Stabilities are not detected and suppressed. S 15.9.A, Rev 0: Thermal II.9 Stability-related instrumentation Conforms Reactor trip signals prevent violation of CHF 4.4.7 aulic Stability Review functionality should be limits before flow instabilities can develop. 15.9.A onsibilities demonstrated by analysis. No unique monitoring is required to detect 7.2 Conformance with Regulatory Criteria hydraulic instabilities. S 15.9.A, Rev 0: Thermal II.10 Ensure plant is free from other Conforms None. 4.4.4 aulic Stability Review instability modes that could violate 4.4.7 onsibilities SAFDLs 15.9.A S 15.9.A, Rev 0: Thermal II.11 D&S System extremely high Partially Conforms RTS system is used instead of a D&S. RTS 4.4.7 aulic Stability Review probability of functioning in the occurs prior to conditions that could initiate 4.4.6 onsibilities event of an AOO. instabilities.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status S 16.0, Rev. 0: Technical All (No Acceptance Criteria for Technical Partially Conforms This DSRS section and its acceptance criteria Ch 16 ifications Number) Specifications is applicable but much of the specific language refers to existing LWR technical specifications or to plant-specific technical specifications to be developed by a COL applicant. For the latter, the DCA contains COL information items, as appropriate, that describe the required development of plant-specific technical specifications that is deferred to the COL applicant referencing the NuScale design. Notwithstanding the above, pursuant to 10 CFR 52.47(a)(11) and consistent with DSRS 16.0, the DCA contains proposed technical specifications that are prepared in accordance with 10 CFR 50.36 and 10 CFR 50.36a. The improved standard technical specification guidance for LWRs specified in this DSRS - NUREGs-1430 through -1434, and NUREG-2194 - were utilized to the extent appropriate and practicable. Additionally, the Technical Specifications Task Force Writers Guide for Plant-Specific Improved Technical Specifications, TSTF-GG-05-01, Revision 1, August 2010 was used to draft the Conformance with Regulatory Criteria specifications. There are a number of technical and editorial differences between the NuScale proposed technical specifications and those presented in the improved standard technical specifications. Consistent with this DSRS 16.0, technical justification for such differences is provided.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 16.1, Rev 1: Risk-Informed II.1 Traditional Engineering Guidelines Partially Conforms This guidance is for revisions being made to 16.1.1 sion Making: Technical existing technical specifications (TS), ifications presumably including deviation from generic or any applicable standard TS. The discussion provided was considered in the development of the NuScale TS, however the specific content is not applicable to development of new generic TS as a part of a DCA. 16.1, Rev 1: Risk-Informed II.2 Probabilistic Guidelines Partially Conforms This guidance applies to revisions being 16.1.1 sion Making: Technical made to existing TS, including deviation ifications from generic or applicable standard TS. The discussion provided was considered in the development of the NuScale TS, however the specific content is not applicable to development of new generic TS as a part of a DCA. 17.1, Rev 2: Quality All Various Not Applicable This guidance is applicable only to existing Not Applicable rance During the Design NRC-approved QA Programs that are based Construction Phases on ANSI N45.2 and its daughter standards. The NuScale QA Program Description (QAPD) is based on NQA-1-2008 and the NQA-1a-2009 addenda, as endorsed in RG 1.28, Rev 4. Since the issuance of SRP Section 17.1, the NRC has issued SRP Conformance with Regulatory Criteria Section 17.5 (based on NQA-1) for the review of QAPDs for new reactor applicants

                                                                                               - including applicants for design certification - under 10 CFR 52. Accordingly, SRP Section 17.5 (rather than SRP Section 17.1) is the appropriate guidance to be applied to the NuScale QAPD.

17.2, Rev 2: Quality All Various Not Applicable This guidance is applicable only to existing Not Applicable rance During the NRC-approved operational QA Programs rations Phase that are based on ANSI N45.2 and its daughter standards.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 17.3, Rev 0: Quality All Various Not Applicable This guidance is applicable only to existing Not Applicable rance Program NRC-approved QA Programs. Since the ription issuance of this SRP section, the NRC has issued SRP Section 17.5 for the review of QAPDs for new reactor applicants - including applicants for design certification

                                                                                                 - under 10 CFR 52. Accordingly, SRP Section 17.5 (rather than SRP Section 17.3) is the appropriate guidance to be applied to the QAPD incorporated into the DCA.

17.4, Rev 1: Reliability II.A Design Certification Conforms None. 17.4 rance Program (RAP) 17.4, Rev 1: Reliability II.B COL Applicant Not Applicable This acceptance criterion is applicable only Not Applicable rance Program (RAP) to COL applicants. 17.5, Rev 1: Quality II.A Organization Partially Conforms The onsite, offsite, operational, and 17.5 rance Program maintenance organizational elements of ription - Design Item II.A.3 are the responsibility of the COL fication, Early Site Permit applicant referencing the certified design. New COL applicants 17.5, Rev 1: Quality II.B Quality Assurance Program Partially Conforms The provisions for site-specific and 17.5 rance Program operational phase of the quality assurance ription - Design program are not applicable to the NuScale fication, Early Site Permit QA program to be applied during the New COL applicants design certification phase, and are to be addressed within the operational QA Conformance with Regulatory Criteria program developed and maintained by the COL applicant referencing the certified design. 17.5, Rev 1: Quality II.C Design Control and Verification Conforms None. 17.5 rance Program ription - Design fication, Early Site Permit New COL applicants

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 17.5, Rev 1: Quality II.D Procurement Document Control Conforms None. 17.5 rance Program ription - Design fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.E Instructions, Procedures, and Conforms None. 17.5 rance Program Drawings (Controlled Documents) ription - Design fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.F Document Control Partially Conforms The site-specific and operational provisions 17.5 rance Program of document control are the responsibility ription - Design of the COL applicant referencing the fication, Early Site Permit certified design. New COL applicants 17.5, Rev 1: Quality II.G Control of Purchased Material, Conforms None. 17.5 rance Program Equipment, and Services ription - Design fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.H Identification and Control of Not Applicable This acceptance criterion governs activities Not Applicable rance Program Materials, Parts, and Components that are the responsibility of the COL ription - Design applicant referencing the certified design. fication, Early Site Permit New COL applicants Conformance with Regulatory Criteria 17.5, Rev 1: Quality II.I Control of Special Processes Not Applicable This acceptance criterion governs activities Not Applicable rance Program that are the responsibility of the COL ription - Design applicant referencing the certified design. fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.J Inspection Partially Conforms The provisions specific to inservice, 17.5 rance Program modification, etc. are the responsibility of ription - Design the COL applicant referencing the certified fication, Early Site Permit design. New COL applicants

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 17.5, Rev 1: Quality II.K Test Control Conforms None. 17.5 rance Program ription - Design fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.L Control of Measuring and Test Conforms None. 17.5 rance Program Equipment ription - Design fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.M Handling, Storage, and Shipping Not Applicable This acceptance criterion governs activities Not Applicable rance Program that are the responsibility of the COL ription - Design applicant referencing the certified design. fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.N Inspection, Test, and Operating Not Applicable This acceptance criterion governs activities Not Applicable rance Program Status that are the responsibility of the COL ription - Design applicant referencing the certified design. fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.O Nonconforming Materials, Parts, or Conforms None. 17.5 rance Program Components ription - Design fication, Early Site Permit New COL applicants Conformance with Regulatory Criteria 17.5, Rev 1: Quality II.P Corrective Action Conforms None. 17.5 rance Program ription - Design fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.Q Quality Assurance Records Conforms None. 17.5 rance Program ription - Design fication, Early Site Permit New COL applicants

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 17.5, Rev 1: Quality II.R Audits Conforms None. 17.5 rance Program ription - Design fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.S Training and Qualification Criteria Conforms None. 17.5 rance Program ription - Design fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.T Training and Qualification - Conforms None. 17.5 rance Program Inspection and Test ription - Design fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.U Nonsafety-Related SSC Quality Conforms None. 17.5 rance Program Controls ription - Design fication, Early Site Permit New COL applicants 17.5, Rev 1: Quality II.V Quality Assurance Program Conforms None. 17.5 rance Program Commitments ription - Design fication, Early Site Permit New COL applicants Conformance with Regulatory Criteria 17.6, Rev 2: Maintenance All Various Not Applicable This SRP section and its acceptance criteria Not Applicable govern a site-specific operational program that is the responsibility of the COL applicant. 18.0, Rev 2: Human II.A Review of the HFE Aspects of a New Conforms None. 18.1 thru 18.12 ors Engineering Plant 18.0, Rev 2: Human II.B Review of the HFE Aspects of Control Not Applicable This acceptance criterion is applicable to Not Applicable ors Engineering Room Modifications existing reactor licensees that request NRC approval of control room modifications.

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 18.0, Rev 2: Human II.C Review of the HFE Aspects of Not Applicable This acceptance criterion is applicable to Not Applicable ors Engineering Modifications Affecting Risk existing reactor licensees that request NRC Important Human Actions approval of plant changes that affect important human actions. Appendix 18-A, Rev 0: C.1.B Review Criteria for Phase 1 (Analysis) Conforms This appendix supersedes DI&C ISG-05, 7.1.5 iting Manual Operator Section 3, "Crediting Manual Operator 18.4 ons in Diversity and Actions in Diversity and Defense-in-Depth 18.6 nse-in-Depth (D3) (D3) Analyses. yses Appendix 18-A, Rev 0: C.2.B Review Criteria for Phase 2 Conforms This appendix supersedes DI&C ISG-05, 7.1.5 iting Manual Operator (Preliminary Validation) Section 3, "Crediting Manual Operator 18.4 ons in Diversity and Actions in Diversity and Defense-in-Depth 18.6 nse-in-Depth (D3) (D3) Analyses. yses Appendix 18-A, Rev 0: C.3.B Review Criteria for Phase 3 Conforms This appendix supersedes DI&C ISG-05, 7.1.5 iting Manual Operator (Integrated System Validation) Section 3, "Crediting Manual Operator 18.4 ons in Diversity and Actions in Diversity and Defense-in-Depth 18.6 nse-in-Depth (D3) (D3) Analyses. yses Appendix 18-A, Rev 0: C.4.B Review Criteria for Phase 3 Conforms This appendix supersedes DI&C ISG-05, 7.1.5 iting Manual Operator (Maintaining Long-Term Integrity of Section 3, "Crediting Manual Operator 18.4 ons in Diversity and Credited Manual Actions in the D3 Actions in Diversity and Defense-in-Depth 18.6 nse-in-Depth (D3) Analysis) (D3) Analyses. yses 19.0, Rev 3: Probabilistic All Various Partially Conforms Evaluation of site-specific hazards and PRA 19.0 Conformance with Regulatory Criteria Assessment and Severe update are COL applicant responsibility. 19.1 dent Evaluation for New 19.2 tors 19.1, Rev 3: Determining All Various Not Applicable Applicable to PRAs used by a licensee to Not Applicable Technical Adequacy of support license amendments for an abilistic Risk Assessment operating reactor. isk-Informed License ndment Requests After l Fuel Load

cale Final Safety Analysis Report P or DSRS Section, Rev: AC AC Title/Description Conformance Comments Section Title Status 19.2, Initial Issuance: All Various Not Applicable Applicable to licensees, plant-specific Not Applicable ew of Risk Information proposals for changes to the licensing basis. to Support Permanent tSpecific Changes to the nsing Basis: General ance 19.3, Rev 0: Regulatory All Various Conforms None. 19.3 tment of Non-Safety ems for Passive Advanced t Water Reactors 19.4, Rev 0: Strategies All Various Partially Conforms Applicable with the exception of 19.4 Guidance to Address acceptance criterion II.17 Boiling Water 20

-of-Large Areas of the                                                                  Reactor: Containment Venting and Vessel t Due to Explosions and                                                                 Flooding (Item B.2.e) which is a BWR specific criterion and acceptance criterion II.20 SFP Mitigative Measures. The SFP mitigating measure is not required by NEI 06-12 and includes a statement that this mitigation strategy is not required if the SFP is below grade and cannot be drained. The NuScale SFP is below grade and cannot be drained.

19.5, Rev 0: Adequacy of All Various Conforms None. 19.5 gn features and tional capabilities tified and described for Conformance with Regulatory Criteria standing Aircraft Impacts

cale Final Safety Analysis Report ISG Section/ Title AC AC Title / Description Conformance Comments Section Status OL-ISG-1: Seismic 1 Seismic Issues addressed in this - This section points out to the guidance provided in 3.7 s of High Frequency Interim Staff Guidance Sections 2, 3, 4, and 5. nd Motion OL-ISG-1 2 Ground Motion Definitions Conforms The definitions provided in Section 3.7 are consistent. 3.7 OL-ISG-1 3 Staff Guidance/Position on the Conforms The CSDRS (and CSDRs-HF) is effectively the SSE for the 3.7 Definitions of Safe-Shutdown DCA. The OBE is specified as 1/3 of the CSDRS thus does and Operating-Basis not require any analysis in the DCA. There are COL items Earthquakes, Use of Various for the applicant to ensure the GMRS is enveloped and to Ground Motions, Seismic have a seismic monitoring program with responses Instrumentation and Operating- following an OBE exceedance. Basis Earthquake Exceedance OL-ISG-1 4 Staff Guidance/Position on Conforms The NuScale design includes a high frequency CSDRS. 3.7 Addressing HF Ground Motion Evaluations OL-ISG-1 5 Staff Comments on the Industry Partially Conforms This discusses laboratory analysis of the site-specific soil 2.5 Draft White Paper on Testing of column. The FSAR includes COL items for the applicant to Dynamic Soil Properties for develop site-specific information. Nuclear Power Plant Combined License Applications and Guidance on Information for Review OL-ISG-2: Financial All Various Not Applicable This ISG is applicable to COL applicants. Not Applicable ifications of Applicants ombined License ications Conformance with Regulatory Criteria OL-ISG-3: Probabilistic All Various Not Applicable Guidance concerning the review of PRA information and Not Applicable Assessment Information severe accident assessments submitted to support DC and pport Design COL applications has been incorporated into SRP 19.0, Rev fication and Combined 3. nse Applications OL-ISG-4: Definition of All Various Not Applicable This ISG is applicable to all ESP and COL applicants Not Applicable truction and on Limited requesting authorization to perform limited work activities k Authorizations or considering preconstruction activities.

cale Final Safety Analysis Report ISG Section/ Title AC AC Title / Description Conformance Comments Section Status OL-ISG-5: GALE86 Code All Five paragraphs under heading Not Applicable The NuScale design is similar to large PWRs in the existing Not Applicable alculation of Routine Final Interim Staff Guidance on fleet for effluent release calculations. However, an oactive Releases in Page 3 - Acceptability of GALE86 alternate methodology is necessary because the existing ous and Liquid Effluents PWRGALE code was developed in the 1980s for evaluation pport Design of the large PWR reactors of that time and does not fication and Combined address the NuScale plant design. nse Applications OL-ISG-6: Evaluation Bullets 1 thru Acceptance Criteria - Partially Conforms This guidance refers to Attachment C. The correct 12.3.6 Acceptance Criteria for 6 (p 3 & 4) Compliance with RG 4.21 reference is Attachment B. This guidance is applicable, FR 20.1406 to Support except for the portions that relate to site-specific, gn Certification and operational aspects that are the responsibility of the COL bined License applicant referencing the NuScale design. The aspects of ications this guidance that pertain to design features, facilities, functions, and equipment that are technically relevant to the NuScale standard plant design are applicable to the DCA. OL-ISG-7: Assessment All Normal and Extreme Winter Conforms Section 3.4 identifies parameter specified for the Extreme 2.3 ormal and Extreme Precipitation Events and their and Normal winter precipitation events. These values are 3.4 er Precipitation Loads Resulting Live Roof Loads used in the structural analysis in 3.8. The COL applicant 3.8 he Roofs of Seismic needs to determine site-specific information to compare gory I Structures to the design parameters. That determination is performed in Section 2.3. OL-ISG-8: Necessary para 1 (p4) First paragraph under heading Conforms None. Ch 16 ent of Plant-Specific Final Interim Staff Guidance, nical Specifications specifying identification and timing of resolution of generic Conformance with Regulatory Criteria technical specification COL action items OL-ISG-8 para 2-4 (p. 4 Second, third, and fourth Not Applicable This portion of the ISG is applicable only to COL applicants. Not Applicable

                         & 5)           paragraphs under heading Final Interim Staff Guidance, specifying compliance options for COL applicants OL-ISG-10: Review of     All            Final paragraph on Page 1 -     Partially Conforms This guidance is applicable except for aspects that are         3.9.5 uation to Address                       Review of Adverse Flow Effects                     BWR-specific.

erse Flow Effects in pment Other Than tor Internals

cale Final Safety Analysis Report ISG Section/ Title AC AC Title / Description Conformance Comments Section Status OL-ISG-11: Finalizing All Licensing-Basis Information Partially Conforms This guidance is applicable except for aspects that are nsing-basis Information Freeze Point; Changes That applicable only to COL applicants or early site permit. Should Not be Considered for Deferral OL-ISG-13: NUREG- 1 Failure Mechanism and Partially Conforms Site-specific aspects that are the responsibility of the COL 11.2.3 Standard Review Plan Radioactivity Releases applicant. on 11.2 and Branch nical Position 11-6 ssing the Consequences Accidental Release of oactive Materials from id Waste Tanks for bined License ications Submitted er 10 CFR Part 52 OL-ISG-13 2 Mitigating Design Features Partially Conforms This guidance is applicable except for site-specific aspects 11.2.3 that are the responsibility of the COL applicant. OL-ISG-13 3 Radioactive Source Term Partially Conforms Site-specific aspects are the responsibility of the COL 11.2.3 (Including Attachment A) applicant. OL-ISG-13 4 Calculations of Transport Not Applicable This acceptance criterion governs site-specific calculations Not Applicable Capabilities in Ground Water or that are the responsibility of the COL applicant referencing Surface Water the certified design. OL-ISG-13 5 Exposure Scenarios and Not Applicable This acceptance criterion governs analysis modeling using Not Applicable Acceptance Criteria site-specific hydrogeological data, site characteristics, and radiological analysis; as such, this guidance is the responsibility of the COL applicant referencing the Conformance with Regulatory Criteria certified design. OL-ISG-13 6 SRP Dose Acceptance Criteria Partially Conforms Site-specific aspects are the responsibility of the COL 11.2.3 applicant. OL-ISG-13 7 Specifications on Tank Waste Partially Conforms Site-specific aspects (e.g., development and 11.2.2 Radioactivity Concentration implementation of the ODCM) are the responsibility of the Levels COL applicant. OL-ISG-13 8 Evaluation Findings for Not Applicable This acceptance criterion is explicitly directed towards the Not Applicable Combined License Reviews review of combined license applications.

cale Final Safety Analysis Report ISG Section/ Title AC AC Title / Description Conformance Comments Section Status OL-ISG-14: Assessing All Area of Review; Review Not Applicable As a supplement to SRP Sections 2.4.12 and 2.4.13, this Not Applicable nd Water Flow and Interfaces; Regulatory guidance governs site-specific hydrogeological data, site sport of Accidental Requirements; Onsite characteristics, and radiological analysis aspects that are onuclide Releases Hydrogeological the responsibility of the COL applicant referencing the Characterization; Contamination certified design. Source and Receptor Location; Groundwater Modeling and Pathway Prediction; and Radioactive Consequence Analysis DC/COL-ISG-15: Post- No Num (p4- New Section C.III.4.3 to Replace Not Applicable This guidance is for COL applicants. Not Applicable bined License 11) Section C.III.4.3 of RG 1.206 mitments DC/COL-ISG-15 No Num Anticipated NRC Revisions of Partially Conforms The portions of this guidance that apply to the DCA Ch 1 (p11-23) NUREG0800, SRP Chapter 1.0 include discussion concerning COL action items and COL information items and not using the term COL holder item. COL action items are identified throughout the FSAR. OL-ISG-16: Compliance All - Not Applicable Requirements in 10 CFR 50.54(hh)(2) were moved to Not Applicable 10 CFR 50.54(hh)(2) and 10 CFR 50.155(b)(2).10 CFR 50.54(hh)(2) is not applicable FR 52.80(d) to design certification applicants; however 10 CFR 52.80(d) requires COL applicants to include a description of the equipment upon which mitigating strategies rely to comply with 10 CFR 50.155(b)(2) to maintain or restore core cooling, containment, and SFP cooling capabilities. OL-ISG-17: Ensuring All - Not Applicable This ISG is applicable to the review of seismic design Not Applicable Conformance with Regulatory Criteria rd-Consistent Seismic information submitted to support combined license (COL) t for Site Response and applications. Structure Interaction yses OL-ISG-19: Gas All Various Not Applicable This guidance is applicable only to reactor plant designs Not Applicable mulation Issues in for which operation of emergency core cooling, residual ty Related Systems heat removal, and containment spray systems relies on pumps (i.e. forced circulation). The NuScale emergency core cooling and decay heat removal systems (the NuScale design does not include a containment spray system) operate via natural circulation, and do not require or include pumps.

cale Final Safety Analysis Report ISG Section/ Title AC AC Title / Description Conformance Comments Section Status OL-ISG-20: All Various Not Applicable Guidance concerning the performance of a SMA Not Applicable ementation of a submitted to support DC and COL applications has been abilistic Risk incorporated into SRP 19.0, Rev 3. ssment-Based Seismic in Analysis for New tors OL-ISG-21: Review of All Guidance for Emergency Gas Not Applicable This guidance is applicable only to nuclear power plants Not Applicable ear Power Plant Designs Turbine Generators (Including that use a gas turbine-driven standby emergency AC g a Gas Turbine Driven Attachment 1) power system - in lieu of emergency diesel generators - to dby Emergency supply power to safety-related or risk-significant nating Current Power equipment for operational events and during postulated em accident conditions. The NuScale design uses onsite backup diesel generators instead of gas turbine generators. However, regardless of the type of standby AC generation used in the NuScale design, the onsite standby AC generation source and the onsite AC distribution system it serves are not safety-related, nor are they relied upon to fulfill safety functions during the first 72 hours following a design basis accident. OL-ISG-22: Impact of All Various Not Applicable This ISG is applicable to COL applicants. Not Applicable truction (Under a bined License) of New ear Power Plant Units perating Units at Multi-Sites OL-ISG-24: All Various Conforms Section 2.0 establishes requirements for 2.0 Conformance with Regulatory Criteria ementation of RG 1.221 hurricane wind speed and missile spectra 3.3 esign-Basis Hurricane consistent with guidance in Regulatory 3.5 Hurricane Missiles 1.221, R0. Specific design requirements are established in Sections 3.3.2 and 3.5.1.4. OL-ISG-25: Changes All Various Not Applicable This ISG is applicable to 10 CFR Part 52, COL licensees with Not Applicable ng Construction Under Changes during Construction license condition. 10 of the Code of ral Regulations Part 52 OL-ISG-26: All Various Not Applicable This ISG is applicable to the review of ESP and COL Not Applicable ronmental Issues applications, including those applicants requesting a ciated with New limited work authorization. tors

cale Final Safety Analysis Report ISG Section/ Title AC AC Title / Description Conformance Comments Section Status OL-ISG-27: Specific All Various Not Applicable This ISG is applicable to the review of ESP, LWA, OL, CP, Not Applicable ronmental Guidance for and COL applications for light water SMR reactor t Water Small Modular technologies. tor OL-ISG-28: Assessing All Various Conforms Provides guidance for DC and COL 19.1 echnical Adequacy of applicants to conform to PRA Standard. Advanced Light-Water tor Probabilistic Risk ssment for the Design fication Application and bined License ication al I&C-ISG-01: Cyber 5. Staff Position Not Applicable Not Applicable rity al I&C-ISG-02: Diversity 1 and 2 Adequate Diversity and Manual Not Applicable Digital I&C-ISG-02 is not applicable to the NuScale DCA. Not Applicable Defense-in-Depth (D3) Operator Actions - Staff Position See DSRS 7.1.5 in Table 1.9-3 which provides information (Pages 2 and 3) on the Diversity and Defense-in-Depth review. al I&C-ISG-02 3 BTP 7-19 Position 4 Challenges - Not Applicable Digital I&C-ISG-02 is not applicable to the NuScale DCA. Not Applicable Staff Position (Page 6) See DSRS 7.1.5 in Table 1.9-3 which provides information on the Diversity and Defense-in-Depth review. al I&C-ISG-02 4 Effects of Common Cause Failure Not Applicable Digital I&C-ISG-02 is not applicable to the NuScale DCA. Not Applicable (CCF) - Staff Position (Pages 8 See DSRS 7.1.5 in Table 1.9-3 which provides information and 9) on the Diversity and Defense-in-Depth review. al I&C-ISG-02 6 Echelons of Defense - Staff Not Applicable Digital I&C-ISG-02 is not applicable to the NuScale Not Applicable Position (Page 12) application. See DSRS 7.1.5 in Table 1.9-3 which provides information on Diversity and Defense-in-Depth review. Conformance with Regulatory Criteria al I&C-ISG-02 7 Single Failure - Staff Position Not Applicable Digital I&C-ISG-02 is not applicable to the NuScale Not Applicable (Page 14) application. See DSRS 7.1.5 in Table 1.9-3 which provides information on Diversity and Defense-in-Depth review. al I&C-ISG-03: Risk- 4 Staff Position Not Applicable Digital I&C-ISG-03 is not applicable to the NuScale DCA. Not Applicable med Digital See DSRS 7.0 in Table 1.9-3 which provides an overview of umentation and the I&C review process. rols al I&C-ISG-04: Highly 1 Interdivisional Communications Not Applicable Digital I&C-ISG-04 is not applicable to the NuScale DCA; Not Applicable grated Control Rooms & - Staff Position (Pages 4 through however, it is addressed as part of topical report NuScale al Communication 8) Power, LLC, TR-1015-18653-P-A, "Design of the Highly ems Integrated Protection System Platform Topical Report."

cale Final Safety Analysis Report ISG Section/ Title AC AC Title / Description Conformance Comments Section Status al I&C-ISG-04 2 Command Prioritization - Staff Not Applicable Digital I&C-ISG-04 is not applicable to the NuScale DCA; Not Applicable Position (Pages 8 through 10) however, it is addressed as part of topical report NuScale Power, LLC, TR-1015-18653-A, "Design of the Highly Integrated Protection System Platform Topical Report." al I&C-ISG-04 3 Multidivisional Control and Not Applicable Digital I&C-ISG-04 is not applicable to the NuScale DCA; Not Applicable Display Stations - Staff Position however, it is addressed as part of topical report NuScale (Pages 11 through 16) Power, LLC, TR-1015-18653-A, "Design of the Highly Integrated Protection System Platform Topical Report." al I&C-ISG-04 3.1 Independence and Isolation Not Applicable Digital I&C-ISG-04 is not applicable to the NuScale DCA; Not Applicable (Pages 11 through 13) however, it is addressed as part of topical report NuScale Power, LLC, TR-1015-18653-A, "Design of the Highly Integrated Protection System Platform Topical Report." al I&C-ISG-04 3.2 Human Factors Considerations Not Applicable Digital I&C-ISG-04 is not applicable to the NuScale DCA; Not Applicable (Pages 13 through 15) however, it is addressed as part of topical report NuScale Power, LLC, TR-1015-18653-A, "Design of the Highly Integrated Protection System Platform Topical Report." al I&C-ISG-04 3.3 Diversity and Defense-in-Depth Not Applicable Digital I&C-ISG-04 is not applicable to the NuScale DCA; Not Applicable (D3) Considerations (Page 15) however, it is addressed as part of topical report NuScale Power, LLC, TR-1015-18653-A, "Design of the Highly Integrated Protection System Platform Topical Report." al I&C-ISG-05: Highly 1 Computer-Based Procedures - Partially Conforms This position is applicable except for site-specific 18.7 grated Control Rooms - Staff Position (Pages 3 through operational elements of subtier NUREG-0899 that are the an Factors 7) responsibility of the COL applicant. al I&C-ISG-05 2 Minimum Inventory - Staff Partially Conforms This acceptance criterion is applicable except for the 18.7 Position (Pages 9 through 11) application of certain subtier guidance. al I&C-ISG-05 3 Crediting Manual Operator Not Applicable This acceptance criterion is superseded by Chapter 18 Not Applicable Conformance with Regulatory Criteria Actions in Diversity and Defense- Appendix 18-A. In-Depth (D3) Analyses (Pages 13 through 21) al I&C-ISG-05 3.1.B Phase 1: Analysis - Review Not Applicable This acceptance criterion is superseded by Chapter 18 Not Applicable Criteria (Pages 15 through 16) Appendix 18-A Criterion 1.B. al I&C-ISG-05 3.2.B Phase 2: Preliminary Validation - Not Applicable This acceptance criterion is superseded by Chapter 18 Not Applicable Review Criteria (Page 18) Appendix 18-A Criterion 2.B. al I&C-ISG-05 3.3.B Phase 3: Integrated System Not Applicable This acceptance criterion is superseded by Chapter 18 Not Applicable Validation - Review Criteria Appendix 18-A.Criterion 3.B. (Pages 19 through 20)

cale Final Safety Analysis Report ISG Section/ Title AC AC Title / Description Conformance Comments Section Status al I&C-ISG-05 3.4.B Phase 4: Maintaining Long-Term Not Applicable This acceptance criterion is superseded by Chapter 18 Not Applicable Integrity of Credited Manual Appendix 18-A.Criterion 4.B. Actions in the D3 Analysis - Review Criteria (Page 21) al I&C-ISG-06: Licensing All Various Not Applicable This guidance is for review of requests for licensing basis Not Applicable ess changes from existing licensees to implement digital I&C upgrades. al I&C-ISG-07: Fuel All Various Not Applicable This guidance is for review of proposed measures for Not Applicable e Facilities protecting digital I&C equipment used as items relied on for safety (IROFS) at fuel cycle facilities from unintentional digital events.

/DPR-ISG-01:              All             Various                          Not Applicable   This guidance governs site-specific programmatic and             Not Applicable rgency Planning for                                                                          design aspects of emergency planning that are the ear Power Plants                                                                            responsibility of the COL applicant referencing the NuScale design.
/DPR-ISG-02:              All             Various                          Not Applicable   Applicable to license holder during decommissioning              Not Applicable rgency Planning                                                                              activities.

ption Requests for mmissioning Nuclear er Plants

/DPR-ISG-03: Review of    All             Various                          Not Applicable   Applicable to license holder during decommissioning              Not Applicable rity Exemptions/License                                                                     activities.

ndment Requests for mmissioning Nuclear er Plants ISG-12-01, Rev 1: All Various Not Applicable This ISG is applicable to holders of, and applicants for, Not Applicable Conformance with Regulatory Criteria pliance with Order EA- operating licenses, construction permits, and combined 49 Concerning licenses. ation Strategies ISG-12-03, Rev 1: All Various Not Applicable This ISG is applicable to holders of, and applicants for Not Applicable pliance with Order EA- operating licenses, construction permits, and combined 51 Concerning Spent licenses. Pool monitoring instrumentation that is capable Pool Instrumentation of monitoring and providing indication of beyond design basis events (i.e., instrumentation that can monitor a wide range of spent fuel pool levels) is part of the NuScale design.

cale Final Safety Analysis Report ISG Section/ Title AC AC Title / Description Conformance Comments Section Status ISG-12-04, Draft: All Various Not Applicable This ISG is for response to the March 2012 50.54(f) request Not Applicable orming a Seismic Margin for information letter. DC/COL-ISG-020 remains the NRCs ssment in Response to current guidance for application of an SMA to new March 2012 Request for reactors licensing. mation Letter ISG-12-05, Draft: All Various Not Applicable This ISG is for response to the March 2012 50.54(f) request Not Applicable ormance of an for information letter. grated Assessment for ding ISG-12-06, Draft: All Various Not Applicable This ISG is for response to the March 2012 50.54(f) request Not Applicable orming a Tsunami, for information letter. e, or Seiche Hazard ssment ISG-13-01, Draft: All Various Not Applicable The information in this guidance is site-specific and is the Not Applicable ating Flooding Hazards responsibility of the COL applicant. to Dam Failure ISG-2015-01, Revision 0: All Various Not Applicable This ISG is applicable to BWR licensees with Mark I and Not Applicable pliance with Phase 2 of Mark II containments. r EA-13-109, Order ifying Licenses with rd to Reliable Hardened ainment Vents Capable peration under Severe dent Conditions ISG-2010-01 1 Fuel Assembly Selection Conforms One fuel assembly design is used in the criticality analysis. 9.1.1 ISG-2010-01 2 Depletion Analysis Conforms The analysis does not take credit for burnup. 9.1.1 Conformance with Regulatory Criteria ISG-2010-01 3a Axial Burnup Profile Conforms The analysis does not take credit for burnup. 9.1.1 ISG-2010-01 3b Rack Model Conforms The rack model analysis is appropriate for conditions. 9.1.1 ISG-2010-01 3c Interfaces Conforms The analysis does not take credit for zoning or a loading 9.1.1 pattern. ISG-2010-01 3d Normal Conditions Conforms The analysis considers the presence of an additional 9.1.1 assembly alongside the fuel storage racks. Due to the spacing and the large number of assemblies in the base analysis model, there is no statistically significant increase in reactivity.

cale Final Safety Analysis Report ISG Section/ Title AC AC Title / Description Conformance Comments Section Status ISG-2010-01 3e Accident Conditions Conforms The analysis considers fuel handling accidents, rack 9.1.1 damage consistent with postulated accidents and full boron dilution. All analyses are within the limits established for normal conditions. ISG-2010-01 4a Area of Applicability Conforms The analysis considers area of applicability in the code 9.1.1 validation. ISG-2010-01 4b Trend Analysis Conforms The analysis includes a trend analysis in the code 9.1.1 validation. ISG-2010-01 4c Statistical Treatment Conforms The analysis includes both a bias term and an uncertainty 9.1.1 derived from the code validation. ISG-2010-01 4d Lumped Fission Products Conforms The analysis does not take credit for burnup. 9.1.1 ISG-2010-01 4e Code-to-Code Comparisons Conforms 9.1.1 ISG-2010-01 5a Precedents Conforms The analysis does not rely upon cited precedents. 9.1.1 ISG-2010-01 5b References Conforms Cited references are publicly available and are referenced 9.1.1 in SFP criticality analyses. ISG-2010-01 5c Assumptions Conforms Assumptions used in the analysis are either observably 9.1.1 conservative or are justified in the presentation of the assumption. Conformance with Regulatory Criteria

cale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 4(f)(1)(i) Perform a plant/site-specific probabilistic risk Partially Conforms Design certification will address reliability of core and 19.0 assessment, the aim of which is to seek such containment heat removal systems, with an update 19.1 improvements in the reliability of core and required by COL applicant to reflect site-specific 19.2 containment heat removal systems as are conditions. significant and practical and do not impact excessively on the plant (II.B.8) 4(f)(1)(ii) Perform an evaluation of the proposed auxiliary Not Applicable This rule requires an evaluation of proposed PWR Not Applicable feedwater system (II.E.1.1) auxiliary feedwater (AFW) systems. The NuScale plant design does have an AFW system like a typical LWR. Neither the literal language nor the intent of this rule applies to the NuScale design. 4(f)(1)(iii) Perform an evaluation of the potential for and Not Applicable The NuScale reactor design differs from large PWRs Not Applicable impact of reactor coolant pump seal damage because the NuScale design does not require or following small-break LOCA (II.K.2.16 and include reactor coolant pumps. Rather, the NuScale II.K.3.25) design uses passive natural circulation of the primary coolant, eliminating the need for reactor coolant pumps. 4(f)(1)(iv) Perform an analysis of the probability of a small- Not Applicable This guidance is applicable only to PWRs that are Not Applicable break LOCA caused by a stuck-open power- designed with power-operated pressurizer relief operated relief valve (PORV) (II.K.3.2) valves. The NuScale design does not use power-operated relief valves. 4(f)(1)(v) Perform an evaluation of the safety effectiveness Not Applicable This requirement applies only to BWRs. Not Applicable of providing for separation of high pressure coolant injection and reactor core isolation cooling system initiation levels (II.K.3.13) Conformance with Regulatory Criteria 4(f)(1)(vi) Perform a study to identify practicable system Not Applicable This requirement applies only to BWRs. Regardless, the Not Applicable modifications that would reduce challenges and issue contemplated by this requirement was related to failures of relief valves (II.K.3.16) power-operated relief valves. The NuScale design does not use power-operated relief valves. 4(f)(1)(vii) Perform a feasibility and risk assessment study to Not Applicable This requirement applies only to BWRs. Not Applicable determine the optimum automatic depressurization system design modifications that would eliminate the need for manual activation (II.K.3.18)

cale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 4(f)(1)(viii) Perform a study of the effect on all core-cooling Not Applicable This requirement applies only to BWRs. Not Applicable modes under accident conditions of designing the core spray and low pressure coolant injection systems (II.K.3.21) 4(f)(1)(ix) Perform a study to determine the need for Not Applicable This requirement applies only to BWRs. Not Applicable additional space cooling to ensure reliable long-term operation of the high pressure coolant injection and reactor core isolation cooling systems (II.K.3.24) 4(f)(1)(x) Perform a study to ensure that the Automatic Not Applicable This requirement applies only to BWRs. Not Applicable Depressurization System, valves, accumulators, and associated equipment and instrumentation will be capable of performing their intended functions (II.K.3.28) 4(f)(1)(xi) Provide an evaluation of depressurization Not Applicable This requirement applies only to BWRs. Not Applicable methods (II.K.3.45) 4(f)(1)(xii) Perform an evaluation of alternative hydrogen Not Applicable Pursuant to 10 CFR 52.47(a)(8) and 10 CFR 50.34(f), Not Applicable control systems paragraph (f)(1)(xii) is excluded from the information required to be included in an application for a design certification. 4(f)(2)(i) Provide simulator capability that correctly models Not Applicable Provisions for simulator capability are the Not Applicable the control room and includes the capability to responsibility of the COL applicant referencing the simulate small-break LOCAs(I.A.4.2) certified design. 4(f)(2)(ii) Establish a program to improve plant procedures, Not Applicable The plant procedure improvement program specified Not Applicable with the program scope to include emergency by this requirement (and development of plant procedures, reliability analyses, human factors procedures) is the responsibility of the COL applicant Conformance with Regulatory Criteria engineering, crisis management, operator referencing the certified design. training, and coordination with INPO and other industry efforts (I.C.9) 4(f)(2)(iii) Provide, for Commission review, a control room Conforms None. 18.7 design that reflects state-of-the-art human factor principles prior to committing to fabrication or revision of fabricated control room panels and layouts (I.D.1) 4(f)(2)(iv) Provide a plant safety parameter display console Conforms The NuScale safety display and indication system is 7.1 (I.D.2) integrated into the control room human-system 7.2.13 interface design rather than having a separate console. 18.7.2

cale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 4(f)(2)(v) Provide for automatic indication of the bypassed Conforms None. 7.1 and operable status of safety systems (I.D.3) 7.2.4 7.2.13 4(f)(2)(vi) Provide the capability of high point venting of Departure The venting of noncondensible gases is unnecessary to 5.4.4 noncondensible gases from the reactor coolant ensure long term core cooling capability. system, and other systems that may be required to maintain adequate core cooling. Systems to achieve this capability shall be capable of being operated from the control room and their operation shall not lead to an unacceptable increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment integrity. (II.B.1) 4(f)(2)(vii) Perform radiation and shielding design reviews of Conforms The NuScale design does not contain vital areas, as 12.4 spaces around systems that may, as a result of an defined by NUREG-0737, Item II.B.2, other than the 19.2 accident, contain accident source term areas for initiating combustible gas monitoring, main radioactive materials, and design as necessary to control room and technical support center. Protection permit adequate access (II.B.2) of necessary equipment from radiation is reasonably assured through demonstrating equipment survivability. 4(f)(2)(viii) Provide capability to promptly obtain and Departure The NuScale design does not rely on primary coolant 9.3.2 analyze samples from the reactor coolant system or containment samples to assess the extent of 11.5 and containment that may contain accident potential core damage. The NuScale design relies upon 12.4 source term radioactive materials (II.B.3) radiation monitors under the bioshield and core exit temperature indications for this assessment. The NuScale design supports an exemption from Conformance with Regulatory Criteria 10 CFR 50.34(f)(2)(viii) design criterion for obtaining and analyzing post-accident samples of the reactor coolant system and containment without exceeding prescribed radiation dose limits. 4(f)(2)(ix) Provide a system for hydrogen control that can Not Applicable Pursuant to 10 CFR 52.47(a)(8) and 10 CFR 50.34(f), Not Applicable safely accommodate hydrogen generated by the Paragraph (f)(2)(ix) is excluded from the information equivalent of a 100% fuel-clad metal water required to be included in an application for a design reaction (II.B.8) certification.

cale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 4(f)(2)(x) Provide a test program and associated model Partially Conforms This requirement is applicable to the DCA except for 5.2.2 development, and conduct tests to qualify aspects specifying PORV block valve testing and reactor coolant system relief and safety valves consideration of ATWS conditions in the testing and, for PWRs, PORV block valves (II.D.1) program. The NuScale design does not use power-operated relief valves. The ATWS provision is not technically relevant to the NuScale design. This aspect of the regulation relates to reactor designs that rely on the relief and safety valves to mitigate the consequences of an ATWS event. The NuScale design supports an exemption from 10 CFR 50.62(c)(1) because the NuScale design relies on protection system diversity to prevent an ATWS, rather than design features to mitigate the condition. As a result, the module response to an ATWS is not analyzed in FSAR Section 15.8, such that the performance of the relief and safety valves is not relied upon to meet the ATWS safety criteria. Therefore, consideration of ATWS conditions in the relief and safety valve test program is not necessary to ensure acceptable performance. 4(f)(2)(xi) Provide direct indication of relief and safety valve Conforms None. 5.2 position (open or closed) in the control room 6.3.1 (II.D.3) 7.1 7.2.13 Conformance with Regulatory Criteria

cale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 4(f)(2)(xii) Provide automatic and manual auxiliary Not Applicable The NuScale design does not have an AFW system as Not Applicable feedwater (AFW) system initiation, and provide would be found at a typical LWR. Also, while the DHRS AFW system flow indication in the control room performs some of the functions of an AFW system at a (II.E.1.2) PWR, the NuScale DHRS is designed for NuScale-specific transients and system characteristics, and its actuation and indication is designed accordingly. Specifically with regard to the portion of this requirement specifying control room flow indication, the DHRS operation involves passive natural circulation flow, with flow characteristics that vary with system conditions, which makes DHRS flow a less useful measurement for the NuScale design. Control room indication for system parameters other than DHRS flow are more appropriate to ensure operators have the information necessary to adequately monitor DHRS operation and reactor core cooling. These parameters include DHRS pressure, valve position indication, and reactor coolant system pressure and temperature. 10 CFR 50.34(f)(2)(xii) is not considered applicable to the NuScale DHRS. Because the language and intent of 10 CFR 50.34(f)(1)(ii) do not apply, the requirement is not applicable to the NuScale design. An exemption would be unnecessary because 10 CFR 50.34(f)(1)(ii) only applies to the technically relevant portions of the TMI requirements. 4(f)(2)(xiii) Provide pressurizer heater power supply and Departure The NuScale design equivalent to hot standby 5.4.5 Conformance with Regulatory Criteria associated motive and control power interfaces condition as stated in 10 CFR 50.34(f)(2)(xiii) is hot 8.3.1 sufficient to establish and maintain natural shutdown condition. The NuScale design does not rely 8.3.2 circulation in hot standby conditions with only on pressurizer heaters to establish and maintain onsite power available (II.E.3.1) natural circulation in hot shutdown conditions.

cale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 4(f)(2)(xiv) Provide containment isolation systems that (A) Departure The NuScale design conforms to 50.34(f)(2)(xiv)(A), (B), 5.2.5 4(f)(2)(xiv)(A) ensure all non-essential systems are isolated (C), and (D). NuScale is requesting an exemption to TMI 6.2.4 4(f)(2)(xiv)(B) automatically; (B) ensure each non-essential requirement 10 CFR 50.34(f)(2)(xiv)(E). For 7.1.5 4(f)(2)(xiv)(C) penetration (except instrument lines) have two 50.34(f)(2)(xiv)(D), the high containment pressure 7.2.13 4(f)(2)(xiv)(D) isolation barriers in series; (C) do not result in analytical limit is above the highest allowable 9.3.6 4(f)(2)(xiv)(E) reopening of the containment isolation valves on containment pressure for leak detection operability. 19.2 resetting of the isolation signal; (D) use a Therefore, the set point for initiating containment containment set point pressure for initiating isolation is compatible with normal operation. containment isolation as low as is compatible Additionally, the containment high pressure analytical with normal operation; and (E) include automatic limit is subatmospheric, therefore, any pressure closing on a high radiation signal for all systems setpoint up to and including the analytical limit will that provide a path to the environs (II.E.4.2) prevent a release to the environs. For 50.34(f)(2)(xiv)(E), the NuScale design differs from that of a traditional large water reactor design of a TMI-era vintage because reactor core uncovery, and resulting core damage, cannot occur without reaching the low low pressurizer level containment isolation setpoint. The pressurizer is an integral part of the reactor vessel, located well above the reactor core, and not connected to the reactor core by piping. Design basis events meet their thermal and hydraulic acceptance criteria without reliance on isolating the CES on a high radiation signal. No design basis event results in degraded or damaged core conditions. Section 19.2 analyses demonstrate severe accident conditions, with resultant core damage, also result in generation of Conformance with Regulatory Criteria reliable containment isolation signals, without reliance on isolation on high containment radiation. An in-containment event resulting in core damage or degradation also results in containment isolation on low low pressurizer level and high containment pressure. An event that leads to core damage or degradation also results in containment isolation on low low pressurizer level. These features provide a reliable alternative means to prevent radiological release from the CES to the environs.

cale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 4(f)(2)(xv) Capability for containment purging/venting Not Applicable The NuScale containment vessel is smaller than a Not Applicable designed to minimize the purging time typical containment building, does not contain sub-consistent with as low as reasonably achievable compartments and does not does not require or (ALARA) (II.E.4.4) incorporate a purge or venting system function as contemplated by this requirement. Personnel access during reactor operation is not needed. In addition, the NuScale ECCS design does not include pumps, and does not involve a typical PWR ECCS recirculation mode where ECCS pump performance relies on containment pressure. Thus purge or vent capability as prescribed by 10 CFR 50.34(f)(2)(xv) is neither required nor included in the NuScale design. This requirement is not technically relevant to the NuScale design. 4(f)(2)(xvi) Establish design criterion for the allowable Not Applicable This requirement applies only to Babcock and Wilcox Not Applicable number of actuation cycles of the ECCS and (B&W) designs. Based on NUREG-0933, this reactor protection system with the expected applicability was the result of unique sensitivity that occurrence rates of severe overcooling events B&W reactor designs exhibited to secondary system (II.E.5.1) transients (both undercooling and overcooling events). The NuScale design does not exhibit such sensitivity. 4(f)(2)(xvii) Provide instrumentation to measure, record, and Conforms None. 6.2.1 readout in the control room: (A) containment 7.1.1 pressure, (B) containment water level, (C) 7.2.13 containment hydrogen concentration, (D) 9.3.2 containment radiation intensity (high level), and 11.5 (E) noble gas effluents at all potential, accident 12.3.4 Conformance with Regulatory Criteria release points. Provide for continuous sampling of radioactive iodines and particulates in gaseous effluents from all potential accident release points, and for onsite capability to analyze and measure these samples (II.F.1) 4(f)(2)(xviii) Provide instruments that provide in the control Conforms None. 4.3.2 room an unambiguous indication of inadequate 6.3 core cooling (II.F.2) 7.0.4 7.2.13

cale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 4(f)(2)(xix) Provide instrumentation adequate for Conforms None. 7.1.1 monitoring plant conditions following an 7.1.2 accident that includes core damage (II.F.3) 7.2.13 19.2 4(f)(2)(xx) Provide power supplies for pressurizer relief Departure The requirements of 10 CFR 50.34(f)(2)(xx) for power 5.4.5 valves, block valves, and level indicators (II.G.1) supplies for pressurizer relief valves and block valves 7.2.13 are not technically relevant to the NuScale design. The 8.1.4 NuScale design supports an exemption from the 8.3.1 portions of 10 CFR 50.34(f)(2)(xx) related to pressurizer 8.3.2 level indicators. 4(f)(2)(xxi) Design auxiliary heat removal systems such that Not Applicable This requirement applies only to BWR designs. Not Applicable necessary automatic and manual actions can be taken to ensure proper functioning when the main feedwater system is not operable (II.K.1.22) 4(f)(2)(xxii) Perform a failure modes and effects analysis of Not Applicable This requirement explicitly states its applicability only Not Applicable the integrated control system (ICS) to include to B&W plant designs. This applicability reflects aspects consideration of failures and effects of input and of the B&W ICS design that were identified following output signals to the ICS (II.K.2.9) the TMI incident as design/reliability deficiencies, and are not pertinent to the NuScale design. 4(f)(2)(xxiii) Provide, as part of the reactor protection system, Not Applicable This requirement applies only to B&W plant designs. Not Applicable an anticipatory reactor trip that would be actuated on loss of main feedwater and on turbine trip (II.K.2.10) 4(f)(2)(xxiv) Provide the capability to record reactor vessel Not Applicable This requirement applies only to BWR designs. Not Applicable water level in one location on recorders that meet normal post-accident recording requirements Conformance with Regulatory Criteria (II.K.3.23) 4(f)(2)(xxv) Provide an onsite Technical Support Center and Partially Conforms None. 13.3 onsite Operational Support Center (III.A.1.2) 4(f)(2)(xxvi) Provide for leakage control and detection in the Partially Conforms This requirement is applicable to the DCA to the extent 5.4 design of systems outside containment that it is relevant to the standard plant design. Aspects of 6.3.1 contain (or might contain) accident source term this requirement that are pertinent to testing and 9.3.2 radioactive materials (III.D.1.1) operational programs are the responsibility of the COL 9.3.4 applicant. 4(f)(2)(xxvii) Provide for monitoring of in-plant radiation and Conforms None. 11.5 airborne radioactivity (III.D.3.3) 11.6 12.3.4

cale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 4(f)(2)(xxviii) Evaluate potential pathways for radioactivity and Conforms None. 6.4.1 radiation that may lead to control room 6.4.4 habitability problems under accident conditions 15.0.3 resulting in an accident source term release (III.D.3.4) 4(f)(3)(i) Provide administrative procedures for evaluating Not Applicable This requirement is the responsibility of the COL Not Applicable operating, design, and construction experience applicant. (I.C.5) 4(f)(3)(ii) Ensure that the QA list required by Criterion II in Conforms None. 3.2 Appendix B to 10 CFR 50 includes all SSC 17.4 important to safety (I.F.1) 4(f)(3)(iii) Establish a QA Program based on the specified Partially Conforms This requirement is applicable to the DCA to the extent 17.5 considerations (I.F.2) it is relevant to design activities in support of the DCA. Aspects of this rule specifying QA program requirements for site-specific design and analysis, operational programs, as-built documentation, and construction and installation are the responsibility of the COL applicant. 4(f)(3)(iv) Provide one or more dedicated containment Departure This requirement is not technically relevant to the 6.2 penetrations, equivalent in size to a single 3-foot- NuScale design. This TMI requirement is based on 19.2 diameter opening (II.B.8) traditional large LWR containment designs and the potential, as of the time of the requirement, need for future containment venting systems to accommodate severe accidents. The NuScale containment vessel design differs from a typical LWR containment structure because of its high-pressure capability. A 3-Conformance with Regulatory Criteria foot opening relative to the NuScale containment is unnecessary. As discussed in Section 6.2.1.1.1, the calculated peak containment for design basis events remains less than the CNV internal design pressure. As discussed in Section 19.2.3, peak containment pressures do not challenge containment integrity for any analyzed severe accident progression. (Refer to TR-0716-50424, Section 2.8). 4(f)(3)(v) Preliminary Design Information - Containment Not Applicable Pursuant to 10 CFR 52.47(a)(8) and 10 CFR 50.34(f), Not Applicable Integrity (II.B.8) paragraph (f)(3)(v) is excluded from the information required to be included in an application for a design certification.

cale Final Safety Analysis Report Item Regulation Description / Title Conformance Comments Section Status 4(f)(3)(vi) For plant designs with external hydrogen Not Applicable The NuScale design does not have external hydrogen Not Applicable recombiners, provide redundant dedicated recombiners. containment penetrations (II.E.4.1) 4(f)(3)(vii) Provide a description of the management plan Not Applicable This requirement is applicable only to applicants and Not Applicable for design and construction activities (II.J.3.1) holders of reactor facility licenses. 191 Assessment of Debris Accumulation on PWR Conforms None. 6.3 Sump Performance 18.2.3 193 Boiling Water Reactor Emergency Cooling Water Not Applicable This Issue is specific to boiling water reactors. Not Applicable System (ECCS) Suction Concerns 199 Implications of Updated Probabilistic Seismic Not Applicable This is applicable to currently-operating plants. Not Applicable Hazard Estimates in Central and Eastern U.S. on Existing Plants 204 Flooding of Nuclear Power Plant Sites Following Not Applicable The information governed by this guidance is site- Not Applicable Upstream Dam Failures specific. Conformance with Regulatory Criteria

cale Final Safety Analysis Report Doc ID Title Conformance Comments Section Status eric Letter 88-14 Instrument Air Supply System Problems Affecting Conforms The IAS furnishes both instrument and service air. IAS 9.3.1 Safety-Related Equipment moisture separators and dryer packages ensure that the instrument air supplied is dry in accordance with the quality standards of ANS/ISA S7.3-R1981. eric Letter 88-15 Electric Power Systems - Inadequate Control Over Partially Conforms Portions relevant to the NuScale passive plant design 8.1.4 Design Processes are considered in the design of electrical systems. 8.3.1 8.3.2 eric Letter 91-06 Resolution of Generic Issue A30, Adequacy Of Partially Conforms No safety-related DC systems; however, relevant 8.1.4 Safety-Related DC Power Supplies Pursuant to portions are considered in the design of the non-Class 8.3.2 10 CFR 50.54(f) 1E EDSS. eric Letter 96-01 Testing of Safety-Related Logic Circuits Conforms None. 7.2.2 7.2.15 8.1.4 eric Letter 2006-02 Grid Reliability and the Impact on Plant Risk and Not Applicable The NuScale Power Plant design does not rely on Not Applicable the Operability of Offsite Power offsite power for safety-related or risk-significant functions. Grid stability studies are the responsibility of a COL applicant that references the NuScale design certification. eric Letter 2007-01 Inaccessible or Underground Power Cable Failures Partially Conforms As described in Chapter 8, the electrical power 8.1 that Disable Accident Mitigation Systems or Cause systems do not include power cables that provide 8.2 Plant Transients. power to equipment with risk-significant or 8.3 safety-related functions. The scope of compliance with the issues addressed by GL 2007-01 is limited to power cables within the scope of 10 CFR 50.65. Conformance is achieved for cable monitoring by the Conformance with Regulatory Criteria COL holder applying the guidance of RG 1.218 as discussed in Chapter 8. eric Letter 2008-01 Managing Gas Accumulation in Emergency Core Partially Conforms NuScale has determined that gas accumulation 5.4 Cooling, Decay Heat Removal, and Containment buildup will not impact ECCS under accident Ch 6 Spray Systems conditions. DHRS does not interface with the RCS. It is connected to the secondary system. tin 2007-01 Security Officer Attentiveness Not Applicable Applicable to holders of operating licenses for Not Applicable nuclear power reactors.

cale Final Safety Analysis Report Doc ID Title Conformance Comments Section Status tin 2011-01 Mitigating Strategies Not Applicable Bulletin 2011-01 was addressed to existing Licensees. Not Applicable It required the Licensee to confirm continue compliance with 10 CFR 50.54(hh)(2). The compliance with 10 CFR 50.54(hh)(2) is addressed in Section 20.2. tin 2012-01 Design Vulnerability in Electric Power System Partially Conforms Consideration of this bulletin is demonstrated by the 8.2.3 conformance with SRP BTP 8-9, which is described in Section 8.2.3. Conformance with Regulatory Criteria

cale Final Safety Analysis Report SRMs) Doc ID Title Conformance Comments Section Status 013 Design Requirements Related to the Evolutionary Conforms Addressed through SECY-90-016 and SECY-93-087. See - Advanced Light Water Reactors Table 1.9-8 for further information. 016 Evolutionary Light-Water Reactor (LWR) Partially Conforms This SECY was directed towards evolutionary ALWR 19.1 Certification Issues and Their Relationship to designs. The applicability of certain SECY-90-016 issues 19.2 Current Regulatory Requirements to passive plants was later established in SECY-93-087 and SECY-94-084. As a passive ALWR design, the NuScale design conforms to the passive plant guidance of SECY-93-087 and SECY-94-084, rather than that of SECY-90-016. See Table 1.9-8 for further information. 241 Level of Detail Required for Design Certification Conforms Incorporated into 10 CFR 52 and implementing NRC - under Part 52 guidance documents. 377 Requirements for Design Certification under Conforms Incorporated into 10 CFR 52 and implementing NRC - 10 CFR Part 52 guidance documents. 074 Prototype Decisions for Advanced Reactor Designs Conforms Incorporated into 10 CFR 52 and implementing NRC - guidance documents. 078 Chapter 11 of the Electric Power Research Not Applicable SECY-91-078 pertains to evolutionary ALWR designs Not Applicable Institutes (EPRI's) Requirements Document and and is not directly applicable to passive plant designs. Additional Evolutionary Light WaterReactor (LWR) Certification Issues 178 ITAAC for Design Certifications and Combined Conforms Incorporated into 10 CFR 52 and implementing NRC 14.3.2 Licenses guidance documents. 210 ITAAC Requirements for Design Review and Conforms Incorporated into 10 CFR 52 and implementing NRC - Conformance with Regulatory Criteria Issuance of FDA guidance documents. 229 Severe Accident Mitigation Design Alternatives for Conforms Incorporated into NRC Orders, regulatory guidance, 19.2.6 Certified Standard Designs and pending rulemaking. 262 Resolution of Selected Technical and Severe Conforms Incorporated into NRC Orders, regulatory guidance, - Accident Issues for Evolutionary Light-Water and pending rulemaking. Reactor (LWR) Designs 053 Use of Design Acceptance Criteria During the Conforms Incorporated into NRC Orders, regulatory guidance, 14.3.6 10 CFR Part 52 Design Certification Reviews and pending rulemaking. 092 The Containment Performance Goal, External Conforms Incorporated into NRC Orders, regulatory guidance, - Events Sequences, and the Definition of and pending rulemaking. Containment Failure for Advanced LWRs

cale Final Safety Analysis Report Doc ID Title Conformance Comments Section Status 087 Policy, Technical, and Licensing Issues Pertaining See Table 1.9-8. None. - to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs 084 Policy and Technical Issues Associated with the Partially Conforms Incorporated into 10 CFR 52 and implementing NRC 5.4 Regulatory Treatment of Non-Safety Systems in guidance documents. The NuScale Fire Protection 8.1 Passive Plant Design (RTNSS) System does not contain any RTNSS equipment. 8.2 However, Section C, Safe Shutdown Requirements, of 8.3 the SECY discusses the stable shutdown condition for 8.4 passive ALWR which is applicable to the NuScale 9.2.5 Power Plant. Appendix 9A 15.0.4 19.3 302 Source-Term-Related Technical and Licensing Conforms Incorporated into 10 CFR 52 and implementing NRC - Issues Relating to Evolutionary and Passive Light- guidance documents. Water-Reactor Designs 132 Policy and Technical Issues Associated with Conforms Incorporated into 10 CFR 52 and implementing NRC 8.1 Regulatory Treatment of Non-Safety Systems in guidance documents. 8.2 Passive Plant Designs 8.3 8.4 19.3 128 Policy and Key Technical Issues Pertaining to the Partially Conforms Section IV of this SECY applies. 19.3 Westinghouse AP600 Standardized Passive Reactor Design 038 Performance-Based Framework for Nuclear Power Not Applicable None. 13.3 Plant Emergency Preparedness Oversight Conformance with Regulatory Criteria 088 Proposed Options to Address Lessons-Learned Not Applicable Site-specific requirements. Not Applicable Review of the U.S. Nuclear Regulatory Commissions Force-On-Force Inspection Program in Response to Staff Requirements Memorandum - COMGEA/COMWCO-14-0001

cale Final Safety Analysis Report Advanced Light-Water Reactor Designs" ssue Description Conformance Comments Section Status Use of a Physically-Based Source Term: Incorporation of engineering judgment Conforms None. 15.0.3 and a more realistic source term in design that deviates from the siting 15.10 requirements in 10 CFR 100. Anticipated Transient without SCRAM (ATWS): Position on the current practices Partially Conforms The NuScale design relies on diversity 15.8 and design features to achieve a high degree of protection against an ATWS. within the module protection system (MPS) to reduce the risk associated with ATWS events. Mid-Loop Operation: Position on design features necessary to ensure a high Not Applicable Design does not use external loops Not Applicable degree of reliability of RHR systems in PWR. and no drain down condition for refueling. Station Blackout (SBO): Position on methods to mitigate the effects of a loss of all Not Applicable The relevance of the SECY-90-016 Not Applicable AC power. SBO issue to passive ALWR designs was deferred to and addressed in Section F of SECY-94-084 and SECY-95-132. The NuScale design conforms to the passive plant guidance these documents. Fire Protection: Positions on design configuration and features the fire protection Conforms None. Appendix 9A system and other management schemes to ensure safe shutdown of the reactor. Intersystem LOCA: Position on acceptable design practices and preventative Conforms None. 9.3.4 measures to minimize the probability of an ISLOCA. 19.2.2 Hydrogen Control: Position on acceptable requirements to measure and mitigate Partially Conforms 6.2.5 the effects of hydrogen produced due to a water reaction with zirconium fuel cladding. Conformance with Regulatory Criteria Core Debris Coolability: Acceptability criteria for cooling area and quenching Conforms None. 19.2 ability regarding corium interaction with concrete. High-Pressure Core Melt Ejection: Position on acceptable design features to Conforms None. 19.2.3 prevent the event of a high-pressure core melt ejection. Containment Performance: Position on acceptable conditional containment Conforms None. 19.1 failure probabilities or other analyses to ensure a high degree of protection from 19.2 the containment. Dedicated Containment Vent Penetration: Position for a dedicated vent Conforms None. 19.2.4 penetration to preclude containment failure resulting from a containment over-pressurization event.

cale Final Safety Analysis Report ssue Description Conformance Comments Section Status Equipment Survivability: Position on the applicability of environmental Conforms None. 19.2.3 qualification and quality assurance requirements related to plant features provided only for severe-accident protection. Elimination of Operating-Basis Earthquake: Position on the applicability of the Conforms By setting the OBE to 1/3 of the SSE it 3.7 OBE in design and the possibility of decoupling the OBE and SSE in the design of is decoupled from the design safety systems. evaluation process. In-Service Testing of Pumps and Valves: Position on periodic testing to confirm Conforms None. 3.9.6 operability of safety-related pumps and valves. Industry Codes and Standards: Position on use of recently developed or modified Conforms NuScale use the latest endorsed all design codes and industry standards in ALWR designs that have not been codes and standards or others on reviewed for acceptability by the NRC. case by case basis. Electrical Distribution: Positions originally addressed by SECY-91-078 that Not Applicable The NuScale electrical system design Not Applicable specified that an evolutionary ALWR provide: (1) an alternate power source to conforms to the passive plant nonsafety-related loads, and (2) at least one offsite circuit connected directly to guidance of SECY-94-084, Section G. each redundant safety division with no intervening nonsafety-related buses. Seismic Hazard Curves and Design Parameters: Position on use of proposed Conforms None. 19.1.5 generic bounding seismic hazard curves and performance of seismic PRA. Leak-Before-Break: Position on use of leak-before-break concept. Conforms LBB is applied to the MS and FW lines 3.6.3 inside containment. Classification of Main Steam Lines in BWRs: Position on the staffs defined Not Applicable Applicable to BWRs. Not Applicable approach for seismic classification of the main steam line in both evolutionary and passive BWRs. Tornado Design Basis: Position on the maximum tornado wind speed to be used Partially Conforms The FSAR uses the maximum tornado 3.3 for a design basis tornado. wind speed of 230 mph found in RG 1.76 Revision 1 rather than the Conformance with Regulatory Criteria outdated 300 mph guidance found in SECY-93-087. Containment Bypass: Position on ALWR design against containment bypass. Conforms None. 15.0.3 Specifically, failure of the containment system to channel fission product releases 19.1 through the suppression pool, or the failure of passive containment cooling heat 19.2 exchanger tubes in large pools of water outside containment. Containment Leak Rate Testing: Position on testing duration for Type C leak rate Partially Conforms None. 6.2.6 testing (prior to rule change). Post-Accident Sampling System: Position on the required capability to analyze Departure The NuScale design supports an 9.3.2 dissolved hydrogen, oxygen, and chloride in accordance with applicable exemption from regulations. 10 CFR 50.34(f)(2)(viii).

cale Final Safety Analysis Report ssue Description Conformance Comments Section Status Level of Detail: Position on a design certification submittal with depth of detail Conforms None. All FSAR similar to that in an FSAR. Sections Prototyping: No guidance provided; information only Conforms None. 1.5 ITAAC: Position on providing ITAAC to demonstrate that a nuclear power plant Conforms None. 14.3 referencing a certified design is built and operates consistent with the design certification. Reliability Assurance Program: Position on providing a description of purpose, Conforms None. 17.4 scope, objectives, and implementation of a design reliability assurance program. Site-Specific PRAs and Analyses of External Events: Position on the inclusion of Conforms None. 19.1 external event analysis beyond the design basis that needs to be addressed as part of the plant PRA during the design certification review. Severe Accident Mitigation Design Alternatives (SAMDAs): Position on the Conforms None. 19.2.6 consideration of SAMDA as part of the final design approval/design certification of an advanced reactor. Generic Rulemaking Related to Design Certification: No guidance provided; Not Applicable Information Only. Not Applicable information only. Defense Against Common-Mode Failures in Digital Instrumentation and Control Conforms None. 7.1.5 Systems: Position on the use of defense-in-depth and diversity of instrumentation and control systems as part of the final design approval/design certification of an advanced reactor. Multiple SG Tube Failures: Position on requiring that analysis of multiple SG Tube Conforms None. 15.6 Failures of 2 to 5 SG tubes be included in the application for design certification of 19.1 passive ALWRs. PRA Beyond Design Certification: Position on requiring conversion of the design Conforms None. 19.1 certification PRA into a plant-specific PRA Conformance with Regulatory Criteria Control Room Annunciator (Alarm) Reliability: Position on recommending that Conforms None. 7.2.13 additional requirements for ALWR alarm systems are necessary to minimize the problems experienced by operating nuclear power plants Regulatory Treatment of Active Nonsafety Systems in Passive Designs: Position on Conforms None. 19.3 the proposed staff approach for resolving the regulatory treatment of the active non-safety systems in passive ALWRs. Definition of Passive Failure: Position on the staff redefining some passive failures Conforms None. 15.0.0 of components as active failures (i.e., check valves) to cause valves to be evaluated in a much more stringent manner than in previous licensing review Thermal-Hydraulic Stability of the SBWR Not Applicable BWR requirement. Not Applicable

cale Final Safety Analysis Report ssue Description Conformance Comments Section Status Safe Shutdown Requirements: Position on using non-safety grade active cooling Conforms The provisions of this SECY are met by 3.1.4 systems to bring a reactor to cold shutdown since non-safety RHR systems do not using the nonsafety-related 5.4.3 comply with the guidance of 1.139 or branch technical position 5-1 containment flood and drain system 7.1 to flood the containment to allow cooldown to cold conditions for disconnection and transfer of NPMs. During shutdown and NPM movement, residual and decay heat removal is provided by heat convection and conduction from the reactor to the reactor pool via the RCS, flooded containment, and the RPV and containment vessel walls. Control Room Habitability: Position on appropriate analytical methods (i.e., dose Conforms None. 15.0.3 limits and accident duration) to be used in determining the acceptability criteria for control room habitably in accordance with regulatory standards. Radionuclide Attenuation: Position on fission product removal processes inside Conforms None. 6.5.3 containment by natural effects and holdup by the secondary building and piping 15.0.3 systems in addition to commission position on containment spray systems for passive ALWRs. Simplification of Offsite Emergency Planning: Position on simplifying off-site Conforms None. 13.3 emergency planning of passive designs due to the estimated low probability of core damage of such designs. Role of the Passive Plant Control Room Operator: Commission position on Conforms None. 18.7 sufficient man-in-the-loop testing and evaluation to be performed and that a fully 18.10 Conformance with Regulatory Criteria functional integrated control room prototype is necessary for passive plant control room designs to demonstrate that functions and tasks are integrated properly into the man/machine interface decisions.

Item 1.10-1: A COL applicant that references the NuScale Power Plant design certification will evaluate the potential hazards resulting from construction activities of the new NuScale facility to the safety-related and risk significant structures, systems, and components of existing operating unit(s) and newly constructed operating unit(s) at the co-located site per 10 CFR 52.79(a)(31). The evaluation will include identification of management and administrative controls necessary to eliminate or mitigate the consequences of potential hazards and demonstration that the limiting conditions for operation of an operating unit would not be exceeded. This COL item is not applicable for construction activities (build-out of the facility) at an individual NuScale Power Plant with operating NuScale Power Modules. 2 1.10-1 Revision 4.1}}