NuScale Standard Plant Design Certification Application
Chapter One Introduction and General Description of the Plant
PART 2 - TIER 2
Revision 2 October 2018 ©2018, NuScale Power LLC. All Rights Reserved COPYRIGHT NOTICE This document bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this document, other than by the U.S. Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC. The NRC is permitted to make the number of copies of the information contained in these reports needed for its internal use in connection with generic and plant-specific reviews and approvals, as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by NuScale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of these reports, the NRC is permitted to make the number of additional copies necessary to provide copies for public viewing in appropriate docket files in public document rooms in Washington, DC, and elsewhere as may be required by NRC regulations. Copies made by the NRC must include this copyright notice in all instances and the proprietary notice if the original was identified as proprietary. NuScale Final Safety Analysis Report Table of Contents
TABLE OF CONTENTS
CHAPTER 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 ...... 1.2-19 1.3 Comparison with Other Facilities ...... 1.3-1 1.4 Identification of Agents and Contractors ...... 1.4-1 1.4.1 Applicant and Program Manager ...... 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
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TABLE OF CONTENTS
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
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LIST OF TABLES
Table 1.1-1: Acronyms and Abbreviations ...... 1.1-4 Table 1.2-1: Overall Characteristics of a NuScale Power Plant ...... 1.2-20 Table 1.2-2: Design Features of a NuScale Power Module ...... 1.2-21 Table 1.3-1: NuScale Plant Comparison with Other Facilities ...... 1.3-2 Table 1.3-2: Safety Systems and Components Required to Protect the Reactor Core - NuScale Comparison with Other Facilities ...... 1.3-3 Table 1.6-1: NuScale Referenced Topical Reports ...... 1.6-2 Table 1.6-2: NuScale Referenced Technical Reports...... 1.6-3 Table 1.7-1: Instrumentation and Controls Functional and Electrical One-Line Diagrams . . . . . 1.7-2 Table 1.7-2: System Drawings ...... 1.7-3 Table 1.8-1: Summary of NuScale Certified Design Interfaces with Remainder of Plant...... 1.8-2 Table 1.8-2: Combined License Information Items ...... 1.8-3 Table 1.9-1: Conformance Status Legend ...... 1.9-4 Table 1.9-2: Conformance with Regulatory Guides ...... 1.9-5 Table 1.9-3: Conformance with NUREG-0800, Standard Review Plan (SRP) and Design Specific Review Standard (DSRS)...... 1.9-46 Table 1.9-4: Conformance with Interim Staff Guidance (ISG) ...... 1.9-195 Table 1.9-5: Conformance with TMI Requirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) ...... 1.9-205 Table 1.9-6: Evaluation of Operating Experience (Generic Letters and Bulletins) ...... 1.9-214 Table 1.9-7: Conformance with Advanced and Evolutionary Light Water Reactor Design Issues (SECYs and Associated SRMs) ...... 1.9-216 Table 1.9-8: Conformance with SECY-93-087, "Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor Designs" ...... 1.9-218
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LIST OF FIGURES
Figure 1.2-1: Conceptual Site Layout...... 1.2-22 Figure 1.2-2: NuScale Functional Boundaries...... 1.2-23 Figure 1.2-3: Schematic of a Single NuScale Power Module and Associated Secondary Equipment...... 1.2-24 Figure 1.2-4: Layout of a Multi-Module NuScale Power Plant...... 1.2-25 Figure 1.2-5: Cutaway Illustration of 12 Module Configuration ...... 1.2-26 Figure 1.2-6: Cutaway View of NuScale Power Module...... 1.2-27 Figure 1.2-7: Steam Generator and Reactor Flow ...... 1.2-28 Figure 1.2-8: Decay Heat Removal System ...... 1.2-29 Figure 1.2-9: Emergency Core Cooling System ...... 1.2-30 Figure 1.2-10: Reactor Building 24’-0” Elevation ...... 1.2-31 Figure 1.2-11: Reactor Building 35'-8" Elevation ...... 1.2-32 Figure 1.2-12: Reactor Building 50'-0" Elevation ...... 1.2-33 Figure 1.2-13: Reactor Building 62'-0" Elevation ...... 1.2-34 Figure 1.2-14: Reactor Building 75'-0" Elevation ...... 1.2-35 Figure 1.2-15: Reactor Building 86'-0" Elevation ...... 1.2-36 Figure 1.2-16: Reactor Building 100'-0" Elevation ...... 1.2-37 Figure 1.2-17: Reactor Building 126'-0" Elevation ...... 1.2-38 Figure 1.2-18: Reactor Building 145'-6" Elevation ...... 1.2-39 Figure 1.2-19: Reactor Building East and West Section View...... 1.2-40 Figure 1.2-20: Reactor Building South Section View...... 1.2-41 Figure 1.2-21: Control Building 50'-0" Elevation...... 1.2-42 Figure 1.2-22: Control Building 63'-3" Elevation...... 1.2-43 Figure 1.2-23: Control Building 76'-6" Elevation...... 1.2-44 Figure 1.2-24: Control Building 100'-0" Elevation ...... 1.2-45 Figure 1.2-25: Control Building 120'-0" Elevation ...... 1.2-46 Figure 1.2-26: Control Building North Section View ...... 1.2-47 Figure 1.2-27: Control Building West Section View...... 1.2-48 Figure 1.2-28: Radioactive Waste Building 71'-0" Elevation...... 1.2-49 Figure 1.2-29: Radioactive Waste Building 82'-0" Elevation...... 1.2-50 Figure 1.2-30: Radioactive Waste Building 100'-0" Elevation...... 1.2-51 Figure 1.2-31: Radioactive Waste Building 120’-0” Elevation ...... 1.2-52
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LIST OF FIGURES
Figure 1.2-32: Radioactive Waste Building North and South Section Views ...... 1.2-53 Figure 1.2-33: Radioactive Waste Building West Section View ...... 1.2-54 Figure 1.7-1a: Electrical Symbols...... 1.7-4 Figure 1.7-1b: Electrical Symbols...... 1.7-5 Figure 1.7-2: Instrumentation and Controls Symbol Legend ...... 1.7-6 Figure 1.7-3a: Piping and Instrumentation Diagram Legends ...... 1.7-7 Figure 1.7-3b: Piping and Instrumentation Diagram Legends ...... 1.7-8 Figure 1.7-3c: Piping and Instrumentation Diagram Legends ...... 1.7-9 Figure 1.7-3d: Piping and Instrumentation Diagram Legends ...... 1.7-10 Figure 1.7-3e: Piping and Instrumentation Diagram Legends ...... 1.7-11 Figure 1.7-3f: Piping and Instrumentation Diagram Legends ...... 1.7-12
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CHAPTER 1 INTRODUCTION AND GENERAL DESCRIPTION OF THE PLANT
1.1 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".
1.1.1 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.
COL Item 1.1-1: A COL applicant that references the NuScale Power Plant design certification will identify the site-specific plant location.
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1.1.2 Containment Type
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 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.
1.1.3 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.
1.1.4 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% to account for measurement uncertainty.
1.1.5 Schedule
COL 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.
1.1.6 Format and Content
1.1.6.1 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.
1.1.6.2 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
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Reports for Nuclear Power Plants: LWR Edition” (SRP). Accordingly, the preparation of this FSAR used the technical guidance provided in the DSRS and SRP as the basis for the NuScale design. A detailed evaluation of conformance with the NuScale DSRS and the SRP is provided in Section 1.9.
1.1.6.3 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.
1.1.6.4 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.
1.1.6.5 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.
1.1.6.6 Acronyms and Abbreviations
A list of acronyms and abbreviations used in this FSAR is provided in Table 1.1-1, Acronyms and Abbreviations.
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Table 1.1-1: Acronyms and Abbreviations Acronym or Description Abbreviation AAC alternate AC power AAPS auxiliary AC power source ABS auxiliary boiler system ABVS Annex Building HVAC system ABWR Advanced Boiling Water Reactor AC alternating current ACI American Concrete Institute ACM Availability Controls Manual ACRS Advisory Committee on Reactor Safeguards AEA Atomic Energy Act AFU air filtration unit AFWS auxiliary feedwater system AHJ authority having jurisdiction AHU air handling unit AIA Authorized Inspection Agency AISC American Institute of Steel Construction AISI American Iron and Steel Institute ALARA as low as reasonably achievable ALU actuation logic unit ALWR advanced light water reactor AMCA Air Movement and Control Association International, Inc. ANB Annex Building ANS American Nuclear Society ANSI American National Standards Institute AO axial offset AOA axial offset anomaly AOO anticipated operational occurrence AOV air-operated valve API American Petroleum Institute APWR Advanced Pressurized Water Reactor AQ augmented quality ARM area radiation monitor ARO all rods out ARS acceleration response spectra ASCE American Society of Civil Engineers ASD adjustable speed drive ASHRAE American Society of Heating, Refrigerating, and Air-Conditioning Engineers ASM American Society for Metals International ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials ATB Administration and Training Building ATWS anticipated transient without scram AVT all-volatile treatment AWS American Welding Society AWWA American Water Works Association BAS boron addition system BAST boric acid storage tank BDBE beyond design basis event BDBEE beyond design basis external event
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation BDG backup diesel generator BOC beginning of cycle BOL beginning of life BOP balance-of-plant BPDS balance-of-plant drain system BPE bioprocessing equipment BPSS backup power supply system BPVC Boiler Pressure Vessel Code BRL Ballistic Research Laboratory BRVS battery room ventilation system BTP Branch Technical Position BWR boiling water reactor CAM continuous air monitor CARS condenser air removal system CAS central alarm station CAS compressed air system CCBE common cause basic event CCDF conditional core damage frequency CCDP conditional core damage probability CCF common cause failure CCFL counter current flow limitation CCFP conditional containment failure probability CDF core damage frequency CDI conceptual design information CDM certified design material CEA control element assembly CES containment evacuation system CET containment event tree CEUS central and eastern United States CFD computational fluid dynamics CFDS containment flooding and drain system CFR Code of Federal Regulations CFT containment flange tool CHF critical heat flux CHFR critical heat flux ratio CFWS condensate and feedwater system CHRS containment heat removal system CHWS chilled water system CILRT containment integrated leak rate test CIM civil interface macro CIP clean-in-place CIS containment isolation system CIV containment isolation valve CLRF conditional large release frequency CLRT containment leakage rate testing CMAA Crane Manufacturers Association of America CMS code management software CMTR certified material test report CNTS containment system CNV containment vessel
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation CNVF containment vessel failure COC certificate of compliance COL combined license COLA combined license application COLR core operating limits report COMS communication system CPRS condensate polisher resin regeneration system CPS condensate polishing system CQC complete quadratic combination CRA control rod assembly CRB Control Building CRDM control rod drive mechanism CRDS control rod drive system CRE control room envelope CRHS control room habitability system CRM control rod misoperation CRVS normal control room HVAC system CSA core support assembly CSDRS certified seismic design response spectra CSDRS-HF certified seismic design response spectra - high frequency CSS containment sampling system CST condensate storage tank CTG combustion turbine generator CUB Central Utility Building CVAP Comprehensive Vibration Assessment Program CVCS chemical and volume control system CWS circulating water system D3 diversity and defense in depth DAC design acceptance criteria DAS distributed antenna system DAS diverse actuation system DAW dry active waste DBA design basis accident DBE design basis event DBPB design basis pipe break DBST design basis source term DBT design basis tornado DC direct current DCA Design Certification Application DCD Design Control Document (Note - this is synonymous with FSAR in this document) DCH direct containment heating DCS distributed control system DDC distributed Doppler coefficient DDG dry dock gate DGB Diesel Generator Building DGBVS Diesel Generator Building HVAC system DHRS decay heat removal system DIM display interface module DMA dimethylamine DNB departure from nucleate boiling
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation DNBR departure from nucleate boiling ratio DOE Department of Energy DOT Department of Transportation D-RAP Design Reliability Assurance Program DSRS Design Specific Review Standard DSS digital safety system DSW dry solid waste DTC Doppler temperature coefficient, fuel temperature coefficient, Doppler coefficient DWS demineralized water system EAB exclusion area boundary EAL Emergency Action Level ECCS emergency core cooling system ECL effluent concentration limit EDL equivalent dead load EDMG extensive damage mitigation guidelines EDNS normal DC power system EDSS highly reliable DC power system EDSS-C EDSS-common EDSS-MS EDSS-module-specific EDV engineering design verification EFDS equipment and floor drainage system EFPD effective full-power days EFPY effective full-power years EHVS 13.8 kV and switchyard system EIM equipment interface module ELAP extended loss of AC power ELVS low voltage AC electrical distribution system ELWR evolutionary light water reactor EMC electromagnetic compatibility EMDAP evaluation model development and assessment process EMDM electromagnetic drive mechanism EMI electromagnetic interference EMVS medium voltage AC electrical distribution system EOC end of cycle EOF emergency operations facility EOL end of life EOP emergency operating procedure EPA Environmental Protection Agency EPG emergency procedure guidelines EPRI Electric Power Research Institute EPZ emergency planning zone EQ equipment qualification EQDP equipment qualification data package EQRF equipment qualification record file ERDA Energy Research and Development Administration ERDS emergency response data system ERF emergency response facility ERO Emergency Response Organization ERS equipment requirement specification ESAS emergency safeguards actuation system
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation ESBWR Economic Simplified Boiling Water Reactor ESF engineered safety feature ESFAS engineered safety features actuation system ESL equivalent static load ESP early site permit ETA ethanolamine ETAP Electrical Transient Analyzer Program FA functional analysis FAC flow-accelerated corrosion FAT factory acceptance test FATT fracture appearance transition temperatures FCI fuel-coolant interaction FCU fan coil unit FDA final design approval FDS fire detection system FEM Federation Europeenne de la Manutention FERC Federal Energy Regulatory Commission FFD fitness-for-duty FFT fast Fourier transform FHA fire hazards analysis FHE fuel handling equipment FHM fuel handling machine FIRS foundation input response spectra FIT flow-indicating transmitter FIV flow-induced vibration FLEX diverse and flexible coping strategies (based on NRC’s Fukushima task force recommendations) FLPRA flooding probabilistic risk assessment FMEA failure modes and effects analysis FOAK first-of-a-kind FOM figure of merit FPGA field programmable gate array FPP Fire Protection Program FPRA fire probabilistic risk assessment FPS fire protection system FRA functional requirements analysis FRP fiber-reinforced polymer FSAR Final Safety Analysis Report (Note - this is synonymous with DCD in this document) FSG FLEX support guidelines FSI fluid-structure interaction FSSA fire safe shutdown analysis FSSD fire safe shutdown FV Fussell-Vesely FW feedwater FWB Fire Water Building FWH feedwater heater FWIV feedwater isolation valve FWLB feedwater line break FWRV feedwater regulating valve FWS feedwater system FWTS feedwater treatment system
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation GAC granulated activated charcoal GDC General Design Criteria GLPS grounding and lightning protection system GMRS ground motion response spectra GQA graded quality assurance GRWS gaseous radioactive waste system GSI generic safety issue GTAW gas tungsten arc weld GTS generic technical specifications HAZ heat-affected zone HCLPF high confidence of low probability of failure HCW high-conductivity waste HDP Hardware Development Plan HDPE high-density polyethylene HED human engineering discrepancy HEI Heat Exchanger Institute HELB high-energy line break HEP human error probability HEPA high-efficiency particulate air HFE human factors engineering, human failure events HFEITS human factors engineering issue tracking system HFP hot full power HIC high integrity container HIPS highly integrated protection system HLHE heavy load handling equipment HMI human machine interface HOV hydraulic-operated valve HP high pressure, horsepower HP-FWH high pressure feedwater heater HPM human performance monitoring HPME high pressure melt ejection HPN health physics network HRA human reliability analysis HRS hardware requirement specification HSI human-system interface HVAC heating ventilation and air conditioning HVDS feedwater heater vents and drains system HWM hard-wired module HZP hot zero power I&C instrumentation and controls IAB inadvertent actuation block IAS instrument air system IBC International Building Code ICIS in-core instrumentation system ICS integrated control system ID inside diameter IDD interface design description IE infrequent event, initiating event IEC International Electrotechnical Commission IEEE Institute of Electrical and Electronics Engineers
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation IES Illuminating Engineering Society of North America IET integral effects test IGSCC intergranular stress-corrosion cracking IHA important human action ILRT integrated leak rate testing INL Idaho National Laboratory INPO Institute of Nuclear Power Operations IOTBS inadvertent opening of the turbine bypass system IP implementation plan IP intermediate pressure IP-FWH intermediate pressure feedwater heater ISA integrated safety analysis, Instrument Society of America ISG interim staff guidance ISI in-service inspection ISLH in-service leak and hydro ISLOCA interfacing systems loss-of-coolant accident ISM independent support motion ISO International Organization for Standardization ISRS in-structure response spectra IST in-service testing ISV integrated system validation ITAAC Inspections, Tests, Analyses, and Acceptance Criteria ITM inspection, testing, and maintenance ITP Initial Test Program IVR in-vessel retention JLD Japan Lessons-Learned Directorate LBB leak-before-break LCO limiting condition for operation LCS local control station LCW low-conductivity waste LER Licensee Event Report LHGR linear heat generation rate LLRT local leak rate test LOCA loss-of-coolant accident LOLA loss of large areas LOOP loss of offsite power LP low pressure LP-FWH low pressure feedwater heater LPSD low power and shutdown LPZ low population zone LRA lower riser assembly LRF large release frequency LRVP liquid ring vacuum pump LRW liquid radioactive waste LRWS liquid radioactive waste system, liquid radwaste system LSH level switch, high LSL level switch, low LSSS limiting safety system setting LTC load manual tap changers LTCC long-term core cooling
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation LTOP low temperature overpressure protection LUHS loss of normal access to the ultimate heat sink LWMS liquid waste management system LWR light water reactor MAE module assembly equipment MC main condenser MCC motor control center MCHFR minimum critical heat flux ratio MCR main control room MCS module control system MCYR module critical year MEL master equipment list MEMS meteorological and environmental monitoring system MFW main feedwater MHA maximum hypothetical accident MHS module heatup system MIB monitoring and indication bus MIC microbiologically induced corrosion MIT Massachusetts Institute of Technology MLA module lifting adapter MLD master logic diagram MM multiple, multi-module MMAF multi-module adjustment factor MMI multi-module issue MMPSF multi-module performance shaping factor MMS moment magnitude scale MOC middle of cycle MOV motor-operated valve MPS module protection system MPT main power transformer MPU magnetic speed pickup MSI main steam isolation MSIBV main steam isolation bypass valves MSIV main steam isolation valve MSLB main steam line break MSO multiple spurious operations MSPI mitigating system performance index MSS main steam system MSSV main steam safety valve MTC moderator temperature coefficient MTU metric tons, uranium MWe megawatt electric MWS maintenance workstation MWt megawatt thermal N/A Not Applicable NDE non-destructive examination NDS nitrogen distribution system NDT non-destructive testing NEI Nuclear Energy Institute NERC North American Electric Reliability Corporation
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation NFA new fuel assembly NFE new fuel elevator NFJC new fuel jib crane NFPA National Fire Protection Association NIC network interface controller NIST National Institute of Standards and Technology NIST-1 NuScale Integral System Test Facility NMS neutron monitoring system NOG nuclear overhead and gantry NPM NuScale Power Module NPP NuScale Power Plant NPS nominal pipe size NPSH net positive suction head NRC Nuclear Regulatory Commission NRF nuclear reliability factor NSA neutron source assembly NSAC Nuclear Safety Analysis Center NSSS nuclear steam supply system NTTF Near-Term Task Force OBE operating basis earthquake OCS operational condition sampling OD outside diameter ODC overspeed detection circuit ODCM Offsite Dose Calculation Manual OE operating experience OER operating experience review OHLHS overhead heavy load handling system ORNL Oak Ridge National Laboratory ORPP Operational Radiation Protection Program OSC operational support center OSHA Occupational Safety and Health Administration OSP overspeed protection system OSU Oregon State University P&ID piping and instrumentation diagram PA protected area PA/GA public address/general alarm PACS priority actuation and control system PAM post-accident monitoring PBX private branch exchange PCA primary coolant activity PCCV prestressed concrete containment vessel PCP Process Control Program PCS plant control system PCT peak cladding temperature PCUS pool cleanup system PDC power distribution center PDC principal design criteria PDIL power dependent insertion limit PDIT differential pressure indicating transmitter PFT process feed tank
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation PGA peak ground acceleration PID proportional integral derivative PING particulate, iodine, and noble gas PIRT phenomena identification and ranking table PIT pressure indicating transmitter PLC programmable logic controller PLD pool leakage detection PLDD programmable logic design description PLDP Programmable Logic Development Plan PLDS pool leakage detection system PLHGR peak linear heat generation rate PLM priority logic module PLRS programmable logic requirement specification PLS plant lighting system PLVVP Programmable Logic Verification and Validation Plan PMF probable maximum flood PMP probable maximum precipitation PORV power-operated relief valve POS plant operating state POV power-operated valve PPE personnel protective equipment PPS plant protection system PRA probabilistic risk assessment PRV pressure relief valve PSCIV primary system containment isolation valves PSCS pool surge control system PSD power spectra density PSMS power supply monitoring system PSS process sampling system PST phase separator tank PSTN public switched telephone network PTAC performance and test acceptance criteria band PTS pressurized thermal shock PVC polyvinyl chloride PVMS plant-wide video monitoring system PWHT post-weld heat treatment PWR pressurized water reactor PWS potable water system PWSCC primary water stress-corrosion cracking PZR pressurizer QA quality assurance QAP Quality Assurance Program QAPD Quality Assurance Program Description QPD quadrant power difference QD quick disconnect QPF quadrant power fractions RAI request for additional information RAP Reliability Assurance Program RAW risk achievement worth RBC Reactor Building crane
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation RBCM Reactor Building components RBVS Reactor Building HVAC system RCA radiologically controlled area RCCA rod control cluster assembly RCCWS reactor component cooling water system RCP reactor coolant pump RCPB reactor coolant pressure boundary RCRA Resource Conservation and Recovery Act RCS reactor coolant system RDT reactor drain tank REA rod ejection accident RETS Radiological Effluent Technical Specifications RFI radio frequency interference RFP refueling pool RFT reactor flange tool RG Regulatory Guide RHR residual heat removal RHX regenerative heat exchanger RIS regulatory issue summary RLE review level earthquake RM radiation monitoring RMS fixed area radiation monitoring system RMTS risk-managed technical specifications RO reverse osmosis ROP Reactor Oversight Process RPCS reactor pool cooling system RPI rod position indication RPS reactor protection system RPV reactor pressure vessel RRS required response spectrum RRV reactor recirculation valve RSA remote shutdown area RSR results summary report RSS remote shutdown station RSV reactor safety valve RTB reactor trip breaker RTD resistance temperature detector RTM requirements traceability matrix RTNDT reference temperature for nil-ductility transition RTNSS regulatory treatment of nonsafety systems RTP rated thermal power RTPTS reference temperature, pressurized thermal shock RTS reactor trip system RVI reactor vessel internals RVV reactor vent valve RWB Radioactive Waste Building RWBCR Radioactive Waste Building control room RWBVS Radioactive Waste Building HVAC system RWDS radioactive waste drain system RWMS radioactive waste management system
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation RWSS raw water supply system RXB Reactor Building RXC reactor core S&Q staffing and qualifications SAFDL specified acceptable fuel design limit SAM seismic anchor motion SAMDA severe accident mitigation design alternative SAMG severe accident management guideline SAR Safety Analysis Report SAS secondary alarm station SAS service air system SAT site acceptance testing SBAC smooth bounding analysis curve SBLB subscale boundary layer boiling SBLOCA small-break loss-of-coolant accident SBM scheduling and bypass module SBO station blackout SBVS Security Building HVAC system SC-I Seismic Category I SC-II Seismic Category II SC-III Seismic Category III SCB Security Buildings SCC stress corrosion-cracking SCDF seismic core damage frequency SCR silicon controlled rectifier SCS secondary sampling system SCWS site cooling water system SDB safety data bus SDIS safety display and indication system SDM shutdown margin SDOE secure development and operational environment SDOF single-degree-of-freedom SDP software development process SDS site drainage system SEB Security Building SECS plant security system SECY Secretary of the Commission, Office of the NRC SEI Structural Engineering Institute SEL seismic equipment list SER Safety Evaluation Report SFA spent fuel assembly SFM safety function module SFP spent fuel pool SFPCS spent fuel pool cooling system SFSS spent fuel storage system SG separation group SG steam generator SG strain gauge SGI safeguards information SGS steam generator system
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation SGTF steam generator tube failure SICS safety information and control system SIL software integrity level SLB steam line break SLP site layout plan SM single module SMA seismic margin assessment SMACNA Sheet Metal and Air Conditioning Contractors' National Association SME subject matter expert SMR small modular reactor SMS seismic monitoring system SNL Sandia National Laboratories SNM special nuclear material SOCA security owner controlled area SOV solenoid-operated valve SPAR standardized plant analysis risk SPND self-powered neutron detector SPS security power system SQDP seismic qualification data package SQRF seismic qualification record form SQUG Seismic Qualification Utility Group SR surveillance requirement SREC standard radiological effluent control SRI Stanford Research Institute SRM staff requirements memorandum SRP Standard Review Plan SRSS square root of the sum of the squares SRST spent resin storage tank SRV sump recirculation valve SRWS solid radioactive waste system SSA safe shutdown analysis SSC structures, systems, and components SSCIV secondary system containment isolation valve SSE safe shutdown earthquake SSI soil-structure interaction SSS secondary sampling system SSSI structure-soil-structure interaction SST station service transformer STPA System-Theoretic Process Analysis SUNSI sensitive unclassified non-safeguards information SVM schedule and voting module SWIS service water intake structure SWMS solid waste management system SWV shear wave velocity SWYD switchyard system TA task analysis TAF top of active fuel TBC Turbine Building crane TBS turbine bypass system TBVS Turbine Building HVAC system
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Table 1.1-1: Acronyms and Abbreviations (Continued) Acronym or Description Abbreviation T/C thermocouple TCU temperature control unit TDH total dynamic head TDS total dissolved solids TEDE total effective dose equivalent TGB Turbine Generator Building TGS turbine generator system TGSS turbine gland sealing system THD total harmonic distortion TIHA treatment of important human actions TIT temperature indicating transmitter TLD thermoluminescent dosimeter TLOSS turbine lube oil storage system TMI Three Mile Island TMR triple module redundancy TNDT nil ductility temperature TOC top of concrete TRS test response spectrum TS technical specifications TSC technical support center TSTF Technical Specification Task Force TUF tubular ultrafiltration UAT unit auxiliary transformer UCRW uncontrolled control rod assembly withdrawal at power UCRWS uncontrolled control rod assembly withdrawal from a subcritical or low power or startup condition UDC uniform Doppler coefficient UHS ultimate heat sink UPS uninterruptible power supply URD Utility Requirements Document URS uniform response spectrum URS upper riser assembly/section USGS United States Geological Survey USI unresolved safety issue USM uniform support motion UTC coordinated universal time UWS utility water system V&V verification and validation VDU video display unit VIT vibration indicating transmitter VLA vented lead-acid VRLA valve-regulated lead-acid VRT voltage regulating transformer WDT watchdog timer WMCR waste management control room WSW wet solid waste WTB Waste Treatment Building ZOI Zone of Influence ZPA zero period acceleration
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1.2 General Plant Description
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).
1.2.1 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
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waste processing equipment. The net total output for a NuScale Power Plant with 12 operating NPMs is approximately 570 MWe.
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.
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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.
6) Site Cooling Water System (SCWS): provides cooling water to plant auxiliary systems. Details associated with location and orientation of the cooling towers as well as equipment design and operation are site specific. Additional details of the SCWS are provided in Section 1.2.1.6.
1.2.1.1 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% 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.]]
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1.2.1.1.1 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 • the absence of RPV or containment penetrations below the top of the reactor core • modularization to enable in-shop fabrication of reactor and containment components
1.2.1.1.2 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.
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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 fluid progresses up through the inside of the SG tubes, it is heated, boiled, and superheated to produce high pressure steam for the turbine generator unit.
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.
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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.
The torispherical RPV head supports the control rod drive mechanisms (CRDMs) and includes penetrations ranging from 2” to 8” diameter for pressurizer spray, 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.
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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.
The CNV is maintained at a vacuum under normal operating conditions. The benefits of maintaining a vacuum in the CNV include: • 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.
1.2.1.1.3 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.
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1.2.1.2 Engineered Safety Features and Emergency Systems
1.2.1.2.1 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.
1.2.1.2.2 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.
1.2.1.2.3 Emergency Core Cooling System
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.
1.2.1.2.4 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.
1.2.1.2.5 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.
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1.2.1.2.6 In-service Inspection of Class 2 and 3 Components
The in-service 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 in-service inspection program.
1.2.1.3 Instrumentation, Controls, and Electrical Systems
The I&C architectural design philosophy incorporates clear interconnection interfaces, separation between safety and non-safety systems, and simplification of system functions. The I&C architecture primarily consists of the following systems, which are described in Section 7.0: • module protection system (MPS) provides information from safety-related sensors monitoring temperature, flow, neutron flux, and pressure data on the NSSS. • 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 non-safety 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.
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• 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 non-safety 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.
1.2.1.4 Power Conversion System
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.
1.2.1.5 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.
1.2.1.6 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
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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)
1.2.1.7 Radioactive Waste Management System
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.
1.2.2 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.
1.2.2.1 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
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comprised of reinforced concrete and polyurethane with a stainless steel surface and steel vertical faceplate that extends into the pool. The horizontal slab is bolted to the top of the bay. 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.
Pipe fittings and electrical connections are provided above the reactor pool water level to permit manual connection and disconnection during NPM installation, refueling outages, and during replacement or removal of NPMs.
There is no safety-related equipment on the 125'-0” elevation. With the exception of demineralized water isolation valves, which are located on the 50'-0" elevation, there is no safety-related equipment below the 75'-0” elevation. Table 3.2-1, Classification of Structures, Systems, and Components, provides the location and classification of systems, structures, and components.
1.2.2.1.1 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
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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.
1.2.2.1.2 Refueling Operations
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.
1.2.2.2 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.
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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. However, the tunnel is for electrical equipment rather than personnel travel between the two buildings.
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 manager’s office • reference room • emergency equipment room • lavatories • break room • telecommunication room
A tunnel allows for personnel travel between the CRB and RXB.
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
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• 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.
1.2.2.2.1 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.
Digital control systems are implemented in a manner that provides independence between safety-related protection systems and nonsafety-related control systems. 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. In the event that the MCR becomes uninhabitable, a remote shutdown station in the Reactor Building provides a secondary location for safe shutdown of the reactors.
1.2.2.2.2 Technical Support Center
A TSC is provided, compliant with the design requirements of NUREG-0696. Section 13.3 provides additional information.
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1.2.2.3 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.
1.2.2.4 Major Systems
1.2.2.4.1 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 closed-loop, two-phase natural circulation cooling system. Two trains of decay heat removal equipment are provided, one attached to each SG loop. Each train is 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.
1.2.2.4.2 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.
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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.
1.2.2.4.3 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.
Additionally, during the NPM startup process, the CVCS is used in conjunction with the module heatup system, to add heat to the reactor coolant to establish natural 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.
1.2.2.5 Other Site Structures
1.2.2.5.1 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.
1.2.2.5.2 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
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• [[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.]]
1.2.2.5.3 Security Buildings
The SCBs are non-safety 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. • verify identity and access status as well as search for contraband items. • provide a structure or space to monitor access into areas of the plant as well as monitoring tamper alarm devices.
1.2.2.5.4 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]]
1.2.2.5.5 [[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
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• DGB HVAC system • maintenance area]]
1.2.3 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.
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Table 1.2-1: Overall Characteristics of a NuScale Power Plant Overall Plant Nominal net output 570 MWe* Number of power modules 12 Power Module Number of reactors One Thermal power rating 160 MWth Nominal gross electrical output 50 MWe RCS normal operating pressure 1,850 psia Steam generator number Two Steam generator type Vertical helical tube Steam cycle Rankine-subcritical regenerative with superheat Turbine type 3,600 rpm, condensing, with extraction Reactor Core Fuel UO2 (<4.95% enrichment) Refueling intervals 24 months * Nominal net output is total gross electrical output minus house loads.
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Table 1.2-2: Design Features of a NuScale Power Module NuScale Design Feature Primary Impact Safety Enhancement RCS contained within the RPV No large diameter primary coolant Eliminates postulated large-break LOCA piping spectrum accidents Natural-convection-cooled core No reactor coolant pumps Eliminates reactor coolant pump accidents, shaft breaks, pump seizure, missile generation and pump leaks High 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 RPV 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 Evacuated 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 Low 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 Reactor pool with partially CNV partially immersed in reactor Provides passive long-term cooling (approximately 90%) immersed NPM pool Passive 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
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Tier 2 1.2-22 Revision 2 NuScale Final Safety Analysis Report General Plant Description SWYD SCWS Pumps SCWS Tower CUB AAP Structures not described in the FSAR Security Owner Controlled Area (SOCA) Fence CRB Protected Area (PA) Boundary Double Fence TGB TGB Cooling Tower Cooling Tower NuScale Power Modules Owner Controlled Area boundary NuScale Power Plant boundary (SOCA Fence) NUSCALE POWER MODULES (NPMS) NUSCALE BDG BDG RXB ANB RWB Figure 1.2-2: NuScale Functional Boundaries 1.2-2: Figure CERTIFIED DESIGN Structures within the Protected Area (PA) Boundary Double Fence SEC Security Owner Controlled Area (SOCA) Fence NUSCALE POWER PLANT SEC Warehouse Certified Design Structures or Components Conceptual Design Structures containing Certified Design SSCs Conceptual Design SSCs Admin/Training Owner Controlled Area
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Tier 2 1.2-24 Revision 2 NuScale Final Safety Analysis Report General Plant Description lti-Module NuScale Power Plant {{ Withheld - See Part 9 }} Part 9 - See Withheld {{ Figure 1.2-4: Layout of a Mu a of Layout 1.2-4: Figure
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Figure 1.2-6: Cutaway View of NuScale Power Module
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Figure 1.2-7: Steam Generator and Reactor Flow
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Figure 1.2-8: Decay Heat Removal System
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Figure 1.2-9: Emergency Core Cooling System
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Figure 1.2-10: Reactor Building 24’-0” Elevation
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Figure 1.2-11: Reactor Building 35'-8" Elevation
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Figure 1.2-12: Reactor Building 50'-0" Elevation
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Figure 1.2-13: Reactor Building 62'-0" Elevation
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Figure 1.2-14: Reactor Building 75'-0" Elevation
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Figure 1.2-15: Reactor Building 86'-0" Elevation
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Figure 1.2-16: Reactor Building 100'-0" Elevation
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Figure 1.2-17: Reactor Building 126'-0" Elevation
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Figure 1.2-18: Reactor Building 145'-6" Elevation
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Figure 1.2-19: Reactor Building East and West Section View
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Figure 1.2-20: Reactor Building South Section View
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Tier 2 1.2-42 Revision 2 NuScale Final Safety Analysis Report General Plant Description {{ Withheld - See Part 9 }} Part 9 - See Withheld {{ Figure 1.2-22: Control Building 63'-3" Elevation 63'-3" Building Control 1.2-22: Figure
Tier 2 1.2-43 Revision 2 NuScale Final Safety Analysis Report General Plant Description {{ Withheld - See Part 9 }} Part 9 - See Withheld {{ Figure 1.2-23: Control Building 76'-6" Elevation 76'-6" Building Control 1.2-23: Figure
Tier 2 1.2-44 Revision 2 NuScale Final Safety Analysis Report General Plant Description {{ Withheld - See Part 9 }} Part 9 - See Withheld {{ Figure 1.2-24: Control Building 100'-0" Elevation 100'-0" Building Control 1.2-24: Figure
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Figure 1.2-27: Control Building West Section View
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Figure 1.2-28: Radioactive Waste Building 71'-0" Elevation
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Figure 1.2-29: Radioactive Waste Building 82'-0" Elevation
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Figure 1.2-30: Radioactive Waste Building 100'-0" Elevation
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Figure 1.2-31: Radioactive Waste Building 120’-0” Elevation
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Figure 1.2-32: Radioactive Waste Building North and South Section Views
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Figure 1.2-33: Radioactive Waste Building West Section View
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1.3 Comparison with Other Facilities
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.
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Table 1.3-1: NuScale Plant Comparison with Other Facilities NuScale Plant Parameter or Feature (per NPM) Typical PWR NuScale Nominal gross electrical output (MWe) 1,186 50 Core thermal output (MWt) 3,411 160 Number of fuel assemblies 193 37 Fuel assembly lattice -17x17 17x17 Effective fuel length (ft) 12 6.56 Fuel rods per fuel assembly 264 264 Average linear heat rate (kW/ft) 5.4 2.5 Number of Control Rod Assemblies 53 16 Design life (years) 40 60 Reactor Coolant System Number of heat transfer loops 4 No External Loops Reactor Coolant Pipes (in.) 27.5-31 None Operating pressure (psia) 2,250 1,850 Hot leg temperature (°F) 618 590 Reactor Vessel Vessel inner diameter (in.) 173 107.5 Thermal shielding- and reflector design Neutron pad design Stacked stainless steel reflector blocks In-core instrumentation Bottom mounted Top mounted Steam Generator Number 4 2 Type Vertical U-tube Helical coil Heat transfer area (ft2) 55,000 Approximately 18,000 Number of tubes 5,626 1,380 Reactor Coolant Pumps 4 0 Pressurizer Internal volume (ft3) 1,800 568 Surge nozzle nominal diameter (in.) 14 None Residual Heat Removal Pumps 2 None Containment Type PCCV Steel Pressure Vessel Inner diameter (ft-in.) 140-0 14-2 Height (ft-in.) 205-0 (inner) 75-8.5 (outer) Containment Spray Pumps 2 None High Pressure Safety Injection Pumps 2 None Charging / Safety Injection Pumps 2 None Low Pressure Safety Injection Pumps 2 None Accumulators 4 None I&C System type Analog Digital Emergency Diesel Generators 2 None Turbine Type 1800 rpm, Tandem Compound Six 3,600 rpm, 10 stage with Superheat Flow Emergency Feedwater Pumps 3 None Charging Pumps (CVCS pumps) 2 2 Used for Safety Injection Yes No Volume Control Tank 1 0 Reactor Component Cooling Water Pumps 4 6 total for 12 NPMs
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Table 1.3-2: Safety Systems and Components Required to Protect the Reactor Core - NuScale Comparison with Other Facilities Safety System or Component Typical PWR NuScale Reactor Pressure Vessel X X Containment Vessel X X Reactor Coolant System X X Decay Heat Removal System X X Emergency Core Cooling System X X Control Rod Drive System X X Containment Isolation System X X Ultimate Heat Sink X X Residual Heat Removal System X Safety Injection System X Refueling Water Storage Tank X Condensate Storage Tank X Auxiliary Feedwater System X Emergency Service Water System X Hydrogen Recombiner or Ignition System X Containment Spray System X Reactor Coolant Pumps X Safety-Related Electrical Distribution System X Alternative Off-Site Power X Emergency Diesel Generators X Safety-Related Class 1E Battery System X Anticipated Transient Without Scram (ATWS) System X
Tier 2 1.3-3 Revision 2 NuScale Final Safety Analysis Report Identification of Agents and Contractors
1.4 Identification of Agents and Contractors
1.4.1 Applicant and Program Manager
NuScale Power, LLC (NuScale) was founded in 2007. The NuScale Power Plant design takes advantage of existing design tools and available nuclear fuel options while leveraging the wealth of knowledge developed through more than 50 years of practical application of light-water-cooled pressurized water reactor (PWR) technology.
1.4.2 Division of Responsibility
NuScale has the overall design responsibility for the NuScale certified design including reactor, containment, primary reactor systems and structures (e.g., Reactor Building, Control Building, and Radioactive Waste Building). NuScale maintains headquarters in Portland, Oregon with engineering design offices in Corvallis, Oregon and Charlotte, North Carolina. In addition, NuScale uses major testing facilities in the United States, Canada, Italy, France, and Germany.
1.4.3 Principal Consultants and Other Participants
Fluor Corporation (Fluor) provides the balance of plant design from its Greenville, South Carolina office. Fluor, a major stakeholder of NuScale, has extensive architectural-engineering experience with over 40 years' experience with commercial nuclear projects, providing operating plant support services to 50 United States and international units at 29 locations. Fluor and its nearly 40,000 employees work with governments and clients in diverse industries around the world to design, construct, and maintain complex and challenging capital projects.
COL 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.
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1.5 Requirements for Additional Technical Information
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.
1.5.1 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.
1.5.1.1 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
Tier 2 1.5-1 Revision 2 NuScale Final Safety Analysis Report Requirements for Additional Technical Information
temperatures to determine mixing coefficients and to obtain single-phase and two-phase pressure drop characteristics of the test assembly for a range of bundle 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.
Tier 2 1.5-2 Revision 2 NuScale Final Safety Analysis Report Requirements for Additional Technical Information
1.5.1.2 Critical Heat Flux Testing - NuFuel HTP2™ Fuel Design
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.
1.5.1.3 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.
Tier 2 1.5-3 Revision 2 NuScale Final Safety Analysis Report Requirements for Additional Technical Information
The electrically-heated test focused on secondary-side performance and consisted of three isolated tubes that were instrumented with well-controlled boundary conditions. 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.
1.5.1.4 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.
Tier 2 1.5-4 Revision 2 NuScale Final Safety Analysis Report Requirements for Additional Technical Information
1.5.1.5 NuScale Integral System Test Program
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.
1.5.1.6 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
Tier 2 1.5-5 Revision 2 NuScale Final Safety Analysis Report Requirements for Additional Technical Information
long control rod drive shaft. A proof-of-concept testing program was conducted to demonstrate the validity of these new designs with regard to performance, reliability, and repeatability of each system. Additional testing to determine misalignment limits 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.
1.5.1.7 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 is to be performed at the AREVA Technical Center in Erlangen, Germany, and is configured as an ambient pressure and temperature test. The ambient test configuration is composed of 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 are immersed in the water under ambient conditions with no coolant flow. During the test, the coupled CRA and control rod drive shaft assembly is 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.
Tier 2 1.5-6 Revision 2 NuScale Final Safety Analysis Report Requirements for Additional Technical Information
Test results are used to confirm the operability of the control rod drive shafts for a range of potential component conditions and distortions. Test results are also used to provide CRA drop time and CRA impact velocity at end of drop.
1.5.2 NuScale Test and Inspection Plans
Information on NuScale test and inspection plans related to plant startup testing is provided in Section 14.2.
Tier 2 1.5-7 Revision 2 NuScale Final Safety Analysis Report Material Referenced
1.6 Material Referenced
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.
Tier 2 1.6-1 Revision 2 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-A, Rev 3 NuScale Topical Report: Quality Assurance December 2016 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, Rev 2 Accident Source Term Methodology September 2017 15 TR-0116-20825-P-A, Rev 1 Applicability of AREVA Fuel Methodology for February 2018 4 the NuScale Design TR-0616-48793, Rev 0 Nuclear Analysis Codes and Methods August 2016 4 Qualification TR-0516-49417, Rev 0 Evaluation Methodology for Stability Analysis of July 2016 4 the NuScale Power Module TR-0516-49422, Rev 0 LOCA Evaluation Model December 2016 15 TR-0915-17564, Rev 1 Subchannel Analysis Methodology February 2017 4 TR-0516-49416, Rev 1 Non-LOCA Methodologies August 2017 15 TR-0116-21012, Rev 1 NuScale Power Critical Heat Flux Correlations November 2017 4 TR-0716-50350, Rev 0 Rod Ejection Analysis Methodology December 2016 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 TR-0915-17772, Rev 0 Methodology for Establishing the Technical December 2016 15 Basis for Plume Exposure Emergency Planning Zones at NuScale Small Modular Reactor Plant Site
Tier 2 1.6-2 Revision 2 NuScale Final Safety Analysis Report Material Referenced 6.2 4.2 5.3 3.8 6.2 3.8 20.2 18.5 18.7 18.1 18.1 18.5 13.5 7.0, 7.2 3.8, 6.2 4.3, 5.3 3.7, 3.8, 9.1 3.7, 3.8, 3.7, 3.12, 3B 3.7, 3.12, FSAR Section 11.1, 11.2, 11.3 5.4, 6.2, 6.3, 15.0 5.4, 6.2, 9.5, 13.6, 14.2, 14.3 9.5, 13.6, 14.2, on Allocation Results Summary Reporton Allocation Results Summary 18.3 s Summary Report 18.2 lts Summary Report 18.5 ence Review Result plementation Plan ram (CVAP) Technical Report TR-0716-50439 Technical Report ram (CVAP) 14.2 3.9, Important Human Actions Results Summary Report Human Actions Results Summary Important 18.6 em Interface Design Results Summary Report Design Results em Interface 18.7 ity Assurance and Qualifications Resu Implementation Plan 18.11 Table 1.6-2: NuScale Referenced Technical Reports Technical Referenced NuScale Table 1.6-2: tional Requirements Analysis and Functi Analysis and tional Requirements vanced Sensor Technical Report Technical vanced Sensor 7.1, 7.2 ed Loss of AC Power (ELAP) Event Power (ELAP) Loss of AC ed 20.1 sk Analysis Results Summary Reportsk Analysis 18.4 and Validation Im and Validation to Explosions and Fires Assessment to Short-Term Transient Analysis point Methodology Technical Report point Methodology Leakage Integr Leakage eering Program management Plan Program management eering eering Operating Experi Operating eering d Control Rod Assembly Designs d Control Physical Security Systems n Methodology and Resultsn Methodology ble Gas Control Report Number Title TR-0716-50424TR-0716-50439 Combusti TR-0816-49833Prog Comprehensive Vibration Assessment TR-0816-50796 Rack Analysis Fuel Storage Due Large Areas of Loss TR-0116-20781TR-0316-22048 Fluence Calculatio TR-0416-48929 Ad Nuclear Steam Supply System TR-0516-49084 of NuScale Design TR-0616-49121 Containment Analysis Methodology Set NuScale Instrument RP-0316-17618 of Treatment Factors Engineering Human TR-0816-50797TR-0816-51127 Mitigation Strategies for Extend TR-0916-51299Fuel an NuFuel HTP2 TR-0916-51502 Long-Term Cooling Methodology TR-1015-18177Seismic Analysis Module NuScale Power TR-1016-51669Temperature Limits Methodology Pressure and Module NuScale Power RP-0316-17619 Human-Syst Factors Engineering Human RP-0914-8534 Human Factors Engin RP-0516-49116 Plan ValidationRoom Staffing Results Control RP-0316-17616RP-0316-17617 Ta Factors Engineering Human Staffing Factors Engineering Human TR-1116-51962TR-1116-52065 NuScale Containment RP-0215-10815Report Release Methodology Technical Effluent RP-0316-17614 Concept of Operations RP-0316-17615 Human Factors Engin Func Factors Engineering Human RP-0914-8543 Human Factors Verification RP-0914-8544RP-1215-20253 Design Factors Engineering Human TR-1117-57216 Plan ValidationRoom Staffing Methodology Control TR-0917-56119Technical Guidelines NuScale Generic Pressure Integrity CNV Ultimate
Tier 2 1.6-3 Revision 2 NuScale Final Safety Analysis Report Drawings and Other Detailed Information
1.7 Drawings and Other Detailed Information
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.
1.7.1 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.
COL Item 1.7-1: A COL applicant that references the NuScale Power Plant design certification will provide site-specific diagrams and legends, as applicable.
1.7.2 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.
COL 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.
Tier 2 1.7-1 Revision 2 NuScale Final Safety Analysis Report Drawings and Other Detailed Information
Table 1.7-1: Instrumentation and Controls Functional and Electrical One-Line Diagrams Figure Title Figure 7.0-1 Overall Instrumentation and Controls System Architecture Diagram Figure 7.0-3 Module Protection System Safety Architecture Overview Figure 7.0-4 Separation Group A Communication Architecture Figure 7.0-5 Separation Group-A and Division I Reactor Trip System and Engineered Safety Features Actuation System Communication Architecture Figure 7.0-6 Reactor Trip Breaker Arrangement Figure 7.0-7 Equipment Interface Module Configuration Figure 7.0-8 Equipment Interface Module Output Figure 7.0-9 Pressurizer Trip Breaker Arrangement Figure 7.0-10 Module Protection System Gateway Diagram Figures 7.0-11a and 7.0-11b Module Protection System Power Distribution Figure 7.0-12 Neutron Monitoring System Ex-Core Block Diagram Figure 7.0-13 Plant Protection System Block Diagram Figure 7.0-14 Safety Display and Indication System Boundary Figure 7.0-15 Safety Display and Indication Hub Figure 7.0-16 Display Interface Module Figure 7.0-17 Module Control System Internal Functions and External Interfaces Figure 7.0-20 Plant Control System Internal Functions and External Interfaces Figure 8.3-1 Station Single Line Diagram Figures 8.3-2a and 8.3-2b 13.8kV and Switchyard System Figures 8.3-3a and 8.3-3b Medium Voltage Electrical System Figures 8.3-4a through 8.3-4z Low Voltage Electrical System Figures 8.3-5a and 8.3-5b Backup Power Supply System Figure 8.3-6 Highly Reliable DC Power System (Common) Figures 8.3-7a and 8.3-7b Highly Reliable DC Power System (Module Specific) Figures 8.3-8a through 8.3-8f Normal DC Power System Figure 11.5-2 Process and Effluent Radiation Monitoring System I&C Configuration
Tier 2 1.7-2 Revision 2 NuScale Final Safety Analysis Report Drawings and Other Detailed Information
Table 1.7-2: System Drawings Figure Title Figure 5.1-2 Reactor Coolant System Simplified Diagram Figure 5.4-9 Decay Heat Removal System Simplified Diagram Figure 6.3-1 Emergency Core Cooling System Figure 6.4-1 Control Room Habitability System Simplified Diagram Figure 9.1.3-1 Spent Fuel Pool Cooling System Diagram Figures 9.1.3-2a and 9.1.3-2b Reactor Pool Cooling System Diagram Figure 9.1.3-3 Pool Cleanup System Diagram Figure 9.1.3-4 Pool Surge Cooling System Diagram Figure 9.2.2-1 Schematic of Reactor Component Cooling Water System Figure 9.2.3-1 Demineralized Water System Schematic Figure 9.2.5-2 Ultimate Heat Sink Qualified Makeup Line Figure 9.2.7-1 Site Cooling Water System Schematic Figure 9.2.8-1 Chilled Water System Piping and Instrumentation Drawing Figure 9.2.9-1 Utility Water System Figure 9.3.1-1 Instrument Air and Service Air Figure 9.3.1-2 Nitrogen Distribution System Figure 9.3.3-1 Radioactive Waste Drain System Simplified Configuration Figure 9.3.3-2 Balance-of-Plant Drain System Simplified Configuration Figure 9.3.4-1 Chemical and Volume Control System Simplified Diagram (Chemical and Volume Control System Module 1 with Module Heatup System Subsystem 6A Shown) Figure 9.3.4-2 Boron Addition System Simplified Diagram Figure 9.3.6-1 Containment Evacuation System Figure 9.3.6-2 Containment Flooding and Drain System Figure 9.4.1-1 Control Room Ventilation System Simplified Diagram Figure 9.4.2-1 Reactor Building HVAC System Simplified Diagram Figure 9.4.3-1 Radioactive Waste Building HVAC System Simplified System Diagram Figure 9.4.4-1 Turbine Building HVAC System Simplified Diagram Figure 9.5.1-1 Fire Protection System Water Supplies and Fire Pumps Figure 9.5.1-2 Fire Protection System Yard Fire Main Loop Figure 10.1-1 Power Conversion System Block Flow Diagram Figure 10.1-2 Flow Diagram and Heat Balance Diagram at Rated Power for Steam and Power Conversion System Cycle Figure 10.2-1 Turbine Generator System Piping and Instrumentation Diagram Figure 10.4-1 Condenser Piping and Instrumentation Diagram Figure 10.4-2 Condenser Air Removal System Piping and Instrumentation Diagram Figure 10.4-3 Circulating Water System Piping and Instrumentation Diagram (Typical of 2) Figures 10.4-4a and 10.4-4b Auxiliary Boiler System Piping and Instrumentation Diagram Figures 11.2-1a through 11.2-1j Liquid Radioactive Waste System Diagram Figures 11.3-1a and 11.3-1b Gaseous Radioactive Waste System Diagram Figure 11.4-1 Block Diagram of the Solid Radioactive Waste System Figure 11.4-2a Process Flow Diagram for SRW Wet Solid Waste Figure 11.4-2b Solid Radioactive Waste System
Tier 2 1.7-3 Revision 2 NuScale Final Safety Analysis Report Drawings and Other Detailed Information
Figure 1.7-1a: Electrical Symbols 6. 9. 12. 16. SINGLE-LINE DIAGRAM LEGEND TWO POSITION TRANSFER SWITCH 1. POWER TRANSFORMERS CONNECTIONS CONDUCTORS/CABLES SWITCHING LINE TYPES
3 WINDING POWER TRANSFORMER SEE 9. CONDUCTORS/CABLES BRANCH LINE H - HIGH SIDE WINDING CONNECTION STAB CABLE BUS ATS = AUTO TRANSFER SWITCH X,Y - LOW SIDE WINDINGS COMMON POINT OF NEUTRAL ATS STATIC = STATIC SWITCH OR OTHER DESCRIPTION SEE 9. CONDUCTORS/CABLES POWER BUS MVA/kVA RATING COOLING METHOD(S) BRANCH LINE IMPEDIENCE RATING VOLTAGE LINE POTENTIAL SPARE CONNECTION OR SPECIAL TERMINAL 4.16 KV- 480V 4.16KV, 2000A, 50KA POWER BUS 1500 KVA 2 WINDING POWER TRANSFORMER RATED VOLTAGE CURRENT LINE ON/AN PRIMARY-SECONDARY VOLTAGE LEVEL CONTINUOUS OPERATING CURRENT Z= 5.75% MVA/KVA RATING(S) FUSED SAFETY SWITCH SHORT CIRCUIT SYMMETRICAL INTERRUPT RATING 40A COOLING METHOD(S) EX: BATTERY AND/OR CHARGER RIGHT - AMPERE RATING IMPEDANCE RATING TEST TERMINAL RELAY ACTUATION, ALARM OR INTERLOCK
0-120V 10KVA STARTER OR CONTACTOR 1ɐ VOLTAGE REGULATING TRANSFORMER 3 INSIDE AN ENCLOSURE PRIMARY-SECONDARY VOLTAGE LEVEL LC - LINE CONTACTOR CABLE ID TAG 2000A SAFETY DISCONNECT SWITCH MVA/KVA RATING(S) LC M -MOTOR TOP RIGHT - AMPERE RATING
# OF PHASES TOP LEFT - SIZE CABLE NAME 2. INSTRUMENT TRANSFORMERS NORMALLY OPEN CONTACT TERMINALS ROTARY SWITCH CURRENT TRANSFORMER 17. 200-5A 1 UPPER RIGHT - QTY PROTECTION DEVICES UPPER LEFT - TRANSFORMER RATIO 10. 4 C400 TERMINAL IN STARTER COMPARTMENT LOWER LEFT - ACCURACY CLASS NORMALLY CLOSED CONTACT NUMBER DENOTES TERMINAL NUMBER HIGH VOLTAGE (> 100kV) CIRCUIT BREAKER A MULTIPLE POSITION SELECTOR SWITCH RIGHT - AMPERE RATING 3000 400 -120V 3 POTENTIAL TRANSFORMER N.O. LEFT - N.O. / BLANK = N.C. 6 TERMINAL IN FIELD TERMINAL BOX UPPER RIGHT - QTY REMOVABLE LINK NUMBER DENOTES TERMINAL NUMBER UPPER LEFT - TRANSFORMER RATIO 2 POSITION SELECTOR SWITCH SHUNT POTENTIAL TRANSFORMER 8 TERMINAL IN LOCAL CONTROL PANEL 400 - 120 V 1 DOUBLE SECONDARY MEDIUM VOLTAGE (1kV - 100kV) CIRCUIT BREAKER NUMBER DENOTES TERMINAL NUMBER 400 - 120 V UPPER RIGHT - QTY RIGHT - AMPERE RATING UPPER LEFT - TRANSFORMER RATIO CONNECTION A LEFT - N.O. / BLANK = N.C. 2000A DISCONNECT SWITCH WITH INTERLOCKED 52 (TO BUS, CABLE TERMINAL OR OTHER). BOTTOM - BREAKER DESIGN M
1200 GROUNDING SWITCH N.O. 6A TOP - DEVICE TYPE 25-1 TRANSFORMER WINDING DESIGNATIONS LEFT - AMPERE RATING TERMINAL IN DCS/PLC PANEL 3. NUMBER DENOTES TERMINAL NUMBER
DELTA CONNECTED TRANSFORMER WINDINGS LINES CROSSING (NO CONNECTION)
INSTRUMENTATION 7 TERMINAL IN LOW VOLTAGE A A WYE CONNECTED TRANSFORMER WINDINGS LOW VOLTAGERIGHT - AMPERE(<1kV) ACRATING OR DC CIRCUIT BREAKER MOTOR CONTROL CENTER (MCC) 200 LINE CONTINUATION 200
N.O. METER OR INSTRUMENT (INSIDE) N.O. 13. NUMBER DENOTES TERMINAL NUMBER XX LEFT - N.O. / BLANK = N.C. A (XX - DENOTES ELECTRICAL CONNECTION A - AMMETER BETWEEN SYMBOLS BEARING SAME A F - FREQUENCY METER CONNECTION ID) PHASOR INDICATION SYN - SYNCHROSCOPE C B V - VOLTMETER 5 3 VAR - VAR METER 200A FUSE TERMINAL IN LOW VOLTAGE SWITCH GEAR W - WATT METER GROUNDING STYLE DESIGNATIONS RIGHT - AMPERE RATING NUMBER DENOTES TERMINAL NUMBER 4. OR LEFT - QUANTITY TI - TEMPERATURE INDICATOR CONNECTING DRAWING IDENTIFIER IDENTIFIER CONNECTING DRAWING DMM - DIGITAL MULTIMETER NP12-YY-DYY-E-SL-YYY-SYY XXXXX XXXXX NP12-YY-DYY-E-SL-YYY-SYY .25KVA NEUTRAL GROUNDING TRANSFORMER TO/FROM ZZZ TO/FROM ZZZ BM 19000-120/240V THERMAL OVERLOAD ELEMENT BM - BATTERY MONITOR CONNECTING EQUIPMENT CONNECTING EQUIPMENT MONITORING DEVICE 2 TERMINAL IN MEDIUM VOLTAGE SWITCH GEAR NUMBER DENOTES TERMINAL NUMBER 51 RELAY/DEVICE 7. LOADS 3 G SEE NP-ES-0303-2923 LIGHTNING ARRESTER MISCELLANEOUS WYE CONNECTED TRANSFORMER KV RIGHT - VOLTAGE RATING FOR RELAY/DEVICE TYPES. TOP LEFT - QUANTITY WINDINGS WITH GROUND 20 18. KW ELECTRICAL LOAD WITH KW RATING CONTROL DEVICES SURGE CAPACITOR 14. DELTA WYE TRANSFORMER WITH LOAD NAME RESISTANCE GROUND 11. DC POWER VFD VARIABLE FREQUENCY DRIVE 2 TOP - KW RATING HP MOTOR WITH HORSEPOWER RATING INVERTER ABBREVIATIONS INPUT: 125 VDC DC INPUT VOLTAGE ELECTRICAL HEATER MOTOR NAME OUTPUT: 120/208 VAC, 3 ɐ GROUND SYMBOL AC OUTPUT VOLTAGE, # ɐ LOCAL LOCAL CONTROLLER N.O. - NORMALLY OPEN CLASS BREAK N.C. - NORMALLY CLOSED 8. GENERATORS S NTERLOCKS S - SAFETY 15. I CLASS BREAK QTY - QUANTITY SYNCHRONOUS GENERATOR NS NS - NONSAFETY SWGR - SWITCH GEAR TRANSFORMER LABELS INPUT: 480 VAC, 3 CHARGER 60 HZ ɐ 5. OPERATING FREQUENCY OUTPUT: 125 VDC AC INPUT VOLTAGE, # ɐ 1800RPM OPERATING RPM DC OUTPUT VOLTAGE K K - KEY MPT MAIN POWER TRANSFORMER 90% PF % POWER FACTOR
M M - MECHANICAL UAT UNIT AUXILIARY TRANSFORMER MULTICELL BATTERY 125 VDC VOLTAGE RATING 150 AH E E - ELECTRICAL AMP HOUR RATING SST 60 CELLS STATION SERVICE TRANSFORMER NUMBER OF CELLS
Tier 2 1.7-4 Revision 2 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) 3ɐ 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 1.7-5 Revision 2 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 1.7-6 Revision 2 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 FLAME ARRESTOR REPRESENT A 0.75" NORMALLY CLOSED 1. ARROW INDICATES FAILURE OR UNACTUATED BALL VALVE NEEDLE VALVE FIRE HYDRANT LOOP SEAL CONCENTRIC REDUCER 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 EDUCTOR CAP (BUTT WELD) OR DRAIN BALL VALVE EXPANSION JOINT SLIDE VALVE XXXX = SYSTEM (NOTE 1) TO XXXX EJECTOR ABBREVIATION FREEZE PROOF HOSE VALVE HINGED EXPANSION JOINT SCREWED END STRAIGHT GLOBE VALVE OPEN VENT INLINE SIGHT GLASS THREE-WAY SLIDE VALVE KEY OPERATED VALVE SWIVEL JOINT HOSE CONNECTION OR DRAIN XXXX = SYSTEM ANGLED GLOBE VALVE ALM TO XXXX 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 TWO-WAY ANGLE VALVE BLIND FLANGE DRY PIPE VALVE DRAIN FUNNEL ANGLE BLOWDOWN VALVE 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 IN-LINE SILENCER FOUR-WAY, FOUR-PORTED GATE VALVE / STARTUP STRAINER 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 GATE VALVE, PLUGGED VENT FLANGES WITH PIPE FLANGES HOSE VALVE ELEVATED FIRE MONITOR PLUG VALVE SIPHON VACUUM BREAKER
THREE-WAY PLUG VALVE ANGLED HOSE VALVE REMOTELY OPERATED FIRE SINGLE BLIND WITH PIPE Y-STRAINER FLANGES (NOTE 1) MONITOR CIRCUIT SETTER BALL VALVE, FLANGED FOUR-WAY PLUG VALVE INJECTION ELEMENT (NOTE 1) FOAM MONITOR FILTER AUTOMATIC VENT VALVE BALL VALVE, PLUGGED FLEXIBLE HOSE ECCENTRIC ROTARY DISC VALVE ELEVATED FOAM MONITOR DISTRIBUTOR BREAK POT THREADED PLUG DIAPHRAGM VALVE INLET AIR FILTER ORIFICE PLATE
PINCH VALVE REMOTELY OPERATED FOAM UNION MONITOR RS REMOVABLE SPOOL TEST PORT BELLOWS SEALED VALVE DIELECTRIC UNION SAFETY SHOWER MECHANICAL COUPLING FLOW NOZZLE CHECK VALVE DIELECTRIC FLANGE DRIP LEG CHECK VALVE WITH 3/32 SAFETY SHOWER W/ EYE WASH WALL PENETRATION ORIFICE IN CLAPPER PITOT TUBE AVERAGING ROOF, FLOOR, OR GROUND STOP CHECK VALVE PENETRATION EYE WASH EXHAUST VENT BALL JOINT SWING ELBOW WAFER CHECK VALVE PRE-ACTION SPRINKLER FREE VENT WITH SCREEN TILTING DISC CHECK VALVE SPRAY SPRINKLER RUPTURE DISK FREE VENT WITHOUT SCREEN ANGLE CHECK VALVE WET SPRINKLER SAMPLE COOLER LIFT CHECK VALVE ALM BUTTERFLY VALVE T W/ TAMPER SWITCH SPRAY DESUPERHEATER EXCESS FLOW CHECK VALVE ALM GATE VALVE T W/ TAMPER SWITCH DESUPERHEATER STEM LEAK-OFF VALVE
PACKED BED TRIPLE DUTY VALVE
Tier 2 1.7-7 Revision 2 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 VARIABLE AREA FLOWMETER W/ INTEGRAL MANUAL PNEUMATIC ACTUATOR (A) ADJUSTING VALVE (PRESSURE-BALANCED DIAPHRAGM ACTUATOR) PRESSURE-REDUCING REGULATOR W/ INTEGRAL OUTLET (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 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)
PRESSURE COMPENSATED M ROTARY MOTOR-OPERATED ACTUATOR STRAIGHT-THRU PRESSURE RELIEF VALVE FO FLOW CONTROL RESTRICTION ORIFICE DRILLED IN VALVE PLUG (TAG NUMBER SHALL BE OMITTED IS VALVE IS 2 POS, 4 WAY DIRECTIONAL S SOLENOID ACTUATOR (OPEN-CLOSE OR MODULATING) OTHERWISE IDENTIFIED) 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 SOLENOID OPERATED .. BACKPRESSURE REGULATOR, INTERNAL PRESSURE TAP )$,/72&/26('326,7,21 VACUUM SAFETY ELEMENT / VACUUM RUPTURE DISK 3 POS, 4 WAY, FLOAT CENTER ACTUATOR WITH MANUAL ACTUATED PARTIAL STROKE DIRECTIONAL CONTROL VALVE, TEST DEVICE SOLENOID OPERATED.. )$,//2&.('$6,6 ACTUATOR WITH REMOTE ACTUATED PARTIAL STROKE BACKPRESSURE REGULATOR, EXTERNAL PRESSURE TAP TEMPERATURE REGULATOR W/ FILLED THERMAL SYSTEM TO RESERVOIR S TEST DEVICE PRESSURE INTENSIFIER S ON-OFF SOLENOID ACTUATOR, NON-LATCHING / AUTOMATIC RESET )$,/$6,6'5,)723(1 THREE-WAY TEMPERATURE REGULATOR W/ FILLED THERMAL SYSTEM CYLINDER S ON-OFF SOLENOID ACTUATOR, LATCHING / ON-OFF SOLENOID PRESSURE REDUCING REGULATOR, INTERNAL PRESSURE TAP R ACTUATOR, MANUAL OR REMOTE RESET )$,/$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) TSE SENSING IS INTERNAL) TAPS THERMAL SAFETY ELEMENT, FUSIBLE PLUG OR DISK CLOSED DURING NORMAL OPERATION (ALL VALVES EXCEPT BUTTERFLY VALVES) E ELECTRIC TO PNEUMATIC CONTROL P SIGNAL CONVERTER DIFFERENTIAL PRESSURE REGULATOR, INTERNAL PRESSURE TAPS STEAM TRAP / GENERIC MOISTURE TRAP T XX ACTION E ELECTRIC TO HYDRAULIC CONTROL (NOTE REQUIRED FOR OTHER TRAP TYPES) H SIGNAL CONVERTER NO - NORMALLY OPEN 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 ELECTROHYDRAULIC LINEAR OR E WITH GLOBE VALVE ALL INLINE VALVES ON P&ID ARE H ROTARY ACTUATOR NORMALLY OPEN. EXCEPTIONS MUST BE NOTED ON P&ID.
Tier 2 1.7-8 Revision 2 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 C H HIGH EFFICIENCY HUMIDIFIER COOLING COIL, HEATING COIL,
N3 FLANGED NOZZLE HIGH AHU FILTER CHILLED WATER HOT WATER CW HW N4 FLANGED NOZZLE W/BLIND MEDIUM EFFICIENCY DEHUMIDIFIER H H MED AHU FILTER HEATING COIL, HEAT HEATING COIL, RECOVERY ELECTRIC N5 FLANGED NOZZLE - ANGLED HR EH LOW EFFICIENCY CENTRIFUGAL FAN
HOPPER LOW AHU FILTER D H DIRECT EXPANSION HEATING COIL, CHEMICAL TOTE VERTICAL VESSEL VESSEL JACKETED VESSEL VERTICAL DRUM COIL STEAM X ST CHARCOAL FILTER AHU FAN CHAR REHEAT COIL HIGH-EFFICIENCY PARTICULATE AIR HEPA FILTER PARALLEL BLADE VAV VARIABLE AIR VOLUME BOX - DAMPER W/ REHEAT COIL T TEST SECTION E S FOR FILTERS W/ T MANIFOLD OPPOSED BLADE DAMPER M I MIXING BOX TANK WITH CONE ROOF TANK WITH DOME ROOF TANK WITH INTERNAL FLOATING ROOF TANK WITH DIAPHRAGM CONE BOTTOM TANK X
HEAT EXCHANGERS & HEAT TRANSFER EQUIPMENT S SECURITY BARRIER B
COMPRESSOR F FIRE DAMPER D SLIDE GATE DAMPER EVAPORATOR S WALL WALL D SMOKE DAMPER SHELL & TUBE HEAT EXCHANGER SHELL & TUBE HEAT DOUBLE PIPE HEAT EXCHANGERTUBE 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 TORNADO DAMPER EXHAUST FAN EXHAUST FAN AIR FLOW CEILING MOUNTED AXIAL UPBLAST CONDENSER EVAPORATOR LOUVER HELICAL COIL HEATER EXTERNAL AIR COOLER COOLING TOWER STEAM GENERATOR ELECTRICAL JACKET PLATE & FRAME COOLING TOWER CELL CHILLER HEAT EXCHANGER VARIABLE AIR PROPELLER FAN VAV VOLUME BOX PUMPS COMPRESSORS & BLOWERS FILTERS MISCELLANEOUS
DRIVERS
ADJUSTABLE SPEED MANUAL ASD DRIVE (MECHANICAL) OR CONDENSER EVAPORATOR 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 M ELECTRIC MOTOR VARIABLE FREQUENCY PUMP VFD DRIVE (ELECTRICAL)
PERMANENT MAGNET SKIMMER ELECTRIC MOTOR VARIABLE SPEED COUPLING VERTICAL CAN CENTRIFUGAL METERING TURBINE SCREW PUMP 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 1.7-9 Revision 2 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' INSTRUMENT TAGGING x LOCATED IN OR ON FRONT OF CONTROL OR MAIN PANEL OR CONSOLE )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 XXXX TEMPERATURE ELEMENTS ARE 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 ) FIRST LETTERS SUCCEEDING LETTERS READOUT 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 Y EVENT, STATE, PRESENCE Y-AXIS AUXILIARY DEVICES S SPEED/FREQUENCY SIC SC SCV SI SSH SSL SSHL SIT ST SY SE SV Z POSITION, DIMENSION Z-AXIS, SAFETY DRIVER, ACTUATOR, T TIC TC TCV TI TSH TSL TSHL TIT TT TY TE TP TW TSE TV TEMPERATURE (NOTE ) INSTRUMENTED SYSTEM UNCLASSIFIED FINAL TDTEMPERATURE, 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 1.7-10 Revision 2 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 63(&,),&$7,216/,67(')259$5,286 COL COLOR H2O WATER NIR NEAR INFRARED REF REFRACTOMETER TURB TURBIDITY x ANALYZER SAMPLE LINES (48,30(177<3(6$5(5(&200(1'$7,216 UV ULTRAVIOLET COMB COMBUSTION H2S HYDROGEN SULFIDE N2 NITROGEN RI REFRACTIVE INDEX 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 GENERIC INSTRUMENT CONNECTION TO PROCESS FLOW DO DISSOLVED OXYGEN LC LIQUID CHROMATOGRAPH OP OPACITY TDL TUNABLE DIODE LASER x 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 IA MAY BE REPLACED BY PA (PLANT AIR), NS (NITROGEN), OR GS (ANY DISP DISPLACER MAG MAGNETIC RES RESISTANCE x GAS SUPPLY). IA PRESSURE xINDICATE 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 xINSTRUMENT ELECTRIC POWER SUPPLY. DRF DRAFT SG STRAIN GAUGE ES xINDICATE VOLTAGE AND TYPE AS REQUIRED, E.G. ES-220 VAC xES MAY BE REPLACED BY 24 VDC, 120 VAC, ETC. TEMPERATURE xINSTRUMENT HYDRAULIC POWER SUPPLY. HS BM BI-METAL RTD RESISTANCE TEMP. DETECTOR TCJ THERMOCOUPLE, TYPE J THRM THERMISTOR xINDICATE PRESSURE AS REQUIRED, E.G. HS-70 PSIG. IR INFRARED TC THERMOCOUPLE TCK THERMOCOUPLE, TYPE K TMP THERMOPILE xELECTRONIC OR ELECTRICAL CONTINUOUSLY VARIABLE OR BINARY SIGNAL. RAD RADIATION TCE THERMOCOUPLE, TYPE E TCT THERMOCOUPLE, TYPE T TRAN TRANSISTOR xFUNCTIONAL DIAGRAM BINARY SIGNAL. RP RADIATION PYROMETER xFUNCTIONAL DIAGRAM CONTINUOUSLY VARIABLE SIGNAL. MISCELLANEOUS xELECTRICAL SCHEMATIC LADDER DIAGRAM SIGNAL AND POWER ANNUNCIATION BURNER, COMBUSTION OTHER POSITION RAILS. ALM ALARM FR FLAME ROD CONC CONCENTRIC PB PUSHBUTTON CAP CAPACITANCE xFILLED THERMAL ELEMENT CAPILLARY TUBE. ANN ANNUNCIATOR IGN IGNITER HOA HAND-OFF-AUTO PC PHOTOCELL EC EDDY CURRENT xFILLED SENSING LINE BETWEEN PRESSURE SEAL AND INSTRUMENT. IR TELEVISION L/R LOCAL/REMOTE SMOKE SMOKE IND INDUCTIVE xGUIDED ELECTROMAGNETIC SIGNAL. UV ULTRA VIOLET MOS MAINTENANCE OVERRIDE SWITCH SYNC SYNCHRONIZATION LAS LASER xGUIDED SONIC SIGNAL. MULTI MULTIVARIABLE TDR TIME DELAY RELAY MAG MAGNETIC xFIBER OPTIC SIGNAL. O/L OVERLOAD TEST TEST MECH MECHANICAL xCOMMUNICATION 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 xDCS, 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 Į ALPHA RADIATION ACC ACCELERATION LC LOAD CELL x DCS-TO-DCS, DCS-TO-PLC, PLC-TO-PC, DCS-TO-FIELDBUS, ETC, TOG TOGGLE ȕ BETA RADIATION EC EDDY CURRENT SG STRAIN GAUGE CONNECTIONS. Y GAMMA RADIATION PROX PROXIMITY WS WEIGH SCALE xCOMMUNICATION LINK AND SYSTEM BUS, BETWEEN DEVICES AND n NEUTRON RADIATION VEL VELOCITY FUNCTIONS OF A FIELDBUS SYSTEM. RAD RADIATION ADSORBED DOSE xLINK FROM AND TO "INTELLIGENT" DEVICES. REM ROENTGEN EQUIVALENT MAN xCOMMUNICATION LINK BETWEEN A DEVICE AND A REMOTE CALIBRATION ADJUSTMENT DEVICE OR SYSTEM. LINK FROM AND TO 'SMART' DEVICES EQUIPMENT DESCRIPTIONS (NOTE ) BOUNDARY IDENTIFICATION MISCELLANEOUS IDENTIFICATIONS x AIR COOLER HEAT EXCHANGER PRESSURE VESSEL xMECHANICAL 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 xPRIMARY LINE DESIGN DUTY: BTU/HR SHELL DP/DT: PSIG/°F SIZE: ID X T-T CLASS OR LINE PIPE CLASS OR NUMBER LINE NUMBER SECONDARY LINE CONFIGURATION: SHELL OP/OT: PSIG/°F CAPACITY: GAL REVISION CLOUD x MOTOR NAME PLATE: HP TUBE DP/DT: PSIG/°F AREA A AREA B (ST) (ST) (ST) (ST) (ST) xSTEAM TRACE LINE TUBE OP/OT: PSIG/°F PUMP
COMPRESSOR CAPACITY: GPM (ET) (ET) (ET) (ET) xELECTRICAL TRACE LINE DP/DT: PSIG/°F TANK DUTY CYCLE: % VENDOR PACKAGE REVISION TRIANGLE OP/OT: PSIG/°F DESIGN HEAD: FT // // // // // DP/DT: PSIG/°F DRAWING CONNECTIONS xPNEUMATIC 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 xHYDRAULIC SIGNAL. MOTOR NAME PLATE: HP CAPACITY: GAL I.D. NUMBER PAGE NUMBER HOLD CLOUD CHILLER 2 R xREFRIGERANT FILTERS CAPACITY: TONS AHU/ ACU/ FCU 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 xOTHER 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 SLOPE FAN POWER: kW X 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 1.7-11 Revision 2 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 QUICK-CHANGE ORIFICE PLATE VORTEX SHEDDING FLOWMETER 025(7+$121((/(0(177<3($33($5621 7+('5$:,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 SINGLE ELEMENT SENSING PROBE TEMPERATURE (MAY BE INSTALLED THROUGH TOP OF VESSEL) DISCIPLINE CODE 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 1.7-12 Revision 2 NuScale Final Safety Analysis Report Interfaces with Certified Design
1.8 Interfaces with Certified Design
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.
1.8.1 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.
1.8.2 Departures
COL 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.
Tier 2 1.8-1 Revision 2 NuScale Final Safety Analysis Report Interfaces with Certified Design
Table 1.8-1: Summary of NuScale Certified Design Interfaces with Remainder of Plant System, Structure, or Component Interface FSAR Type Section Turbine Generator Buildings CDI 1.2.2 Annex Building CDI 1.2.2 Cooling towers, pump houses, and associated structures, systems, and CDI 1.2.2, components (e.g., cooling tower basin, circulating water pumps, cooling 10.4.5 tower fans, chemical treatment building, etc.) Security Buildings CDI 1.2.2 Central Utility Building CDI 1.2.2 Diesel Generator Buildings CDI 1.2.2 Offsite power transmission system, main switchyard, and transformer area CDI 8.2 Auxiliary AC power system CDI 8.3.1 Site cooling water system CDI 9.2.7 Circulating water system CDI 10.4.5 Grounding and lightning protection system CDI 8.3.1 Plant exhaust stack CDI 9.4.2 Potable and sanitary water systems COL 9.2.4 Resin tanks for the condensate polishing system COL 10.4 Site drainage system COL N/A Raw water system COL 9.2.9 Site-specific design parameters, geographic and demographic COL Table 2.0-1, 2.1, 2.2, characteristics, meteorological characteristics, nearby industrial, 2.3, 2.4, 2.5, 3.3, 3.4 transportation, and military facilities, hydrologic characteristics, geology, seismology, and geotechnical characteristics, weather conditions and site topography, flooding Site-specific communications COL 9.5.2 Turbine generators COL 3.5-1 Operational Support Center COL 13.3
Tier 2 1.8-2 Revision 2 NuScale Final Safety Analysis Report Interfaces with Certified Design
Table 1.8-2: Combined License Information Items
Item No. Description of COL Information Item Section COL Item 1.1-1: A COL applicant that references the NuScale Power Plant design certification will identify the 1.1 site-specific plant location. COL 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. COL 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. COL 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. COL 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. COL 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. COL 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. COL 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. COL 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 design 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. COL 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. COL 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 potential accident, or provide site-specific design alternatives. COL 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. COL 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. COL 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. COL 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. COL 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. COL 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.
Tier 2 1.8-3 Revision 2 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 3.4-2: 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. COL 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. COL 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. COL 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. COL 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. COL 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. COL Item 3.5-1: A COL applicant that references the NuScale Power Plant certified design will provide a missile 3.5 analysis for the turbine generator which demonstrates that protection from turbine missiles is accomplished by using barriers. COL Item 3.5-2: Not Used. 3.5 COL Item 3.5-3: A COL applicant that references the NuScale Power Plant certified design 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. COL 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 FSAR Tier 2, Section 3.5.1.4. COL 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 CNV and the area under the bioshield, identify the location of high- and moderate-energy lines, and update Table 3.6-1 as necessary. COL 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 CNV (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. COL 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 and design appropriate protection features. This includes an evaluation and disposition of multi-module impacts in common pipe galleries, and evaluations regarding subcompartment pressurization. The COL applicant will update Table 3.6-2 as appropriate.
Tier 2 1.8-4 Revision 2 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 3.6-4: Not used. 3.6 COL 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. COL 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. COL 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. COL 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. COL 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. COL 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. COL 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. COL 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).
Tier 2 1.8-5 Revision 2 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 3.7-9: 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 V/H spectral ratios used in establishing the site-specific foundation input response spectra and performance-based response spectra for the vertical direction. COL 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. 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 east and west wing walls and pool walls. See FSAR Table 3.7.2-32. 5) The site-specific in-structure response spectra for the fuel storage racks will be shown to be bounded by the in-structure response spectra in Figure 3-6 through Figure 3-14 of TR-0816-49833. 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 3.7.2-119 • Control Building south wall at EL 120’-0”: bounded by in-structure response spectra in Figure 3.7.2-121 If not, the standard design will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demands. COL Item 3.7-11: A COL applicant that references the NuScale Power Plant design certification will perform a 3.7 site-specific analysis that, if applicable, 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. COL 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. COL 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. COL 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.
Tier 2 1.8-6 Revision 2 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 3.8-4: 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. COL 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. COL 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. COL 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 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. COL 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. COL 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 ASME OM Code and 10 CFR 50.55a. COL 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 ASME 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). COL Item 3.9-7: Not Used. COL 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. COL 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. COL 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. COL Item 3.9-11: A COL applicant that references the NuScale Power Plant design certification will implement a 3.9 CRDS Operability Assurance Program that meets the requirements described in NUREG-0800, SRP 3.9.4, Revision 3, Acceptance Criteria II.4.
Tier 2 1.8-7 Revision 2 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 3.9-12: 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 CSDRS, the standard design of NPM components will be shown to have appropriate margin or should be appropriately modified to accommodate the site-specific demand. 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 ASME BPVC, 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 3.10-1: 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 in-service inspection program for 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 1.8-8 Revision 2 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 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 RCS 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 1.8-9 Revision 2 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 and accessibility assessment, steam plant corrosion product deposition 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 6.3-1: 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 in-service inspection program plans in accordance with Section XI of the ASME Code, and will establish the implementation milestones for the program. The COL applicant will identify the applicable edition of the ASME Code 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 in-service inspection program for multiple NuScale Power Modules, including any alternative to the code that may be necessary to implement such an in-service inspection program. 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.
Tier 2 1.8-10 Revision 2 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 7.2-1: 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 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 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 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: The design of the switchyard and the connections to an offsite power system are site-specific 8.2 and are the responsibility of the combined license (COL) applicant. A COL applicant that references the NuScale Power Plant design certification will describe the 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. 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.
Tier 2 1.8-11 Revision 2 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 9.1-7: 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 submit an 9.1 evaluation of the spent fuel storage racks that addresses structural, thermal-hydraulic, criticality, and material analysis aspects of the design. This evaluation is dependent on a vendor-specific design and the as-built configuration of spent fuel storage racks. The design of the spent fuel storage racks is considered acceptable when it meets the criteria of Appendix D to Design Specific Review Standard 3.8.4. 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, including an initial test program, a schedule for re-testing these systems, and the actions to be taken for minimizing leakage from such systems. COL Item 9.3-2: A COL applicant that references the NuScale Power Plant design certification will develop the 9.3 post-accident sampling contingency plans for using the process sampling system and the containment evacuation system off-line radiation monitor to obtain reactor coolant and containment atmosphere samples. The contingency plan will describe the process for collecting representative samples and disposing radioactive samples. A COL applicant will identify temporary equipment (e.g., temporary shielding, sample transport cask, etc.) required to support post-accident sampling. 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. 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.
Tier 2 1.8-12 Revision 2 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 9.5-2: 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 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: Not used. 11.2 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. 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 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).
Tier 2 1.8-13 Revision 2 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 11.5-1: 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 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. 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 the main steam and main feedwater piping penetrations through the Reactor Building pool wall between the NuScale Power Module bays and the Reactor Building steam galleries near the 100 ft elevation (Shown on Figure 3.6-16 and Figure 3.6-17). 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.
Tier 2 1.8-14 Revision 2 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 12.5-1: 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 and which complies with the guidance in NUREG-0696, “Functional Criteria for Emergency Response Facilities," NUREG-0737 Supplement 1, "Clarification of TMI Action Plan Requirements - Requirements for Emergency Response Capability," and NSIR/DPR-ISG-01, "Interim Staff Guidance - Emergency Planning for Nuclear Power Plants.” 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 1.8-15 Revision 2 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 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) COL 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. COL 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. COL 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 COL 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.
Tier 2 1.8-16 Revision 2 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 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. COL Item 13.5-6: Not used. 13.5 COL 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 writer’s 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. COL 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. COL 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 COL 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, Rev 0 NuScale Design of Physical Security Systems. COL 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. COL 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 (SSC). COL 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. COL Item 13.6-6: A COL applicant that references the NuScale Power Plant design certification will provide a 13.6 Cyber Security Plan. COL Item 13.7-1: A COL applicant that references the NuScale Power Plant design certification will provide a 13.7 description of the applicant’s 10 CFR 26 compliant fitness-for-duty (FFD) program for plant operations. COL Item 13.7-2: A COL applicant that references the NuScale Power Plant design certification will provide a 13.7 description of the applicant’s 10 CFR 26 compliant fitness-for-duty (FFD) program for construction. COL 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.
Tier 2 1.8-17 Revision 2 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 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. COL 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. COL 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. COL 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. COL 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. COL 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). COL 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. COL 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. COL 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. COL 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. COL 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 plant’s life. COL 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. COL 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. COL 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. COL 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. COL 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. COL 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. COL 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.
Tier 2 1.8-18 Revision 2 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 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. COL 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). COL 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). COL 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). COL 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). COL 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. COL 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 PRA and modify, as necessary, for applicability to the as-built, as-operated PRA. COL 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. COL 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). COL 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.” COL 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. COL Item 20.1-1: A COL applicant that references the NuScale Power Plant design certification will ensure 20.1 equipment and structures credited for diverse and flexible coping strategies are designed to be available following a site-specific seismic hazard. COL Item 20.1-2: A COL applicant that references the NuScale Power Plant design certification will determine if a 20.1 flood hazard is applicable at the site location. If a flood hazard is applicable, then the COL applicant will ensure equipment and structures credited for diverse and flexible coping strategies are designed to be available following a site-specific flood (including wave action) hazard. COL Item 20.1-3: A COL applicant that references the NuScale Power Plant design certification will determine if 20.1 high wind and applicable missile hazards are applicable at the site location. If high wind and applicable missile hazards are applicable, then the COL applicant will ensure equipment and structures credited for diverse and flexible coping strategies are designed to be available following a site-specific high wind and applicable missile hazards. COL Item 20.1-4: A COL applicant that references the NuScale Power Plant design certification will determine if 20.1 snow, ice and extreme cold temperature hazards are applicable at the site location. If snow, ice and extreme cold hazards are applicable, the COL applicant will ensure equipment and structures credited for diverse and flexible coping strategies are designed to be available following a site-specific snow, ice or extreme cold temperature hazard.
Tier 2 1.8-19 Revision 2 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 20.1-5: A COL applicant that references the NuScale Power Plant design certification will determine if 20.1 extreme high temperature hazard is applicable at the site location. If extreme high temperature hazard is applicable, the COL applicant will ensure equipment and structures credited for diverse and flexible coping strategies are designed to be available following a site-specific extreme high temperature hazard. COL Item 20.1-6: Not used. 20.1 COL Item 20.1-7: Not used. 20.1 COL 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. COL Item 20.2-1: A COL applicant that references the NuScale Power Plant design certification will develop 20.2 enhanced firefighting capabilities by implementing the guidance in NRC guidance document “Developing Mitigating Strategies/Guidance for Nuclear Power Plants to Respond to Loss of Large Areas of the Plant in Accordance with B.5.b of the February 25, 2002, Order” dated February 25, 2005 (Reference 20.2-3). 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). COL 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. COL Item 20.3-1: A COL applicant that references the NuScale Power Plant design certification will ensure that the 20.3 severe accident management guidelines, diverse and flexible coping strategies support guidelines (FSGs), and extensive damage mitigation guidelines are integrated with the emergency operating procedures consistent with Recommendation 8.1 of SECY-11-0093, “Near-Term Report and Recommendations for Agency Actions Following the Events in Japan.” COL Item 20.4-1: A COL applicant that references the NuScale Power Plant design certification will perform an 20.4 analysis that demonstrates the emergency response organization staff has the ability to implement the strategies of the emergency operating procedures, severe accident mitigation guidelines, diverse and flexible coping strategies support guidelines (FSGs), and extensive damage mitigation guidelines. The analysis will be performed with the offsite response organization access to onsite being impeded. The event shall be a loss of all onsite and offsite alternating current power and loss of normal access to the ultimate heat sink. COL Item 20.4-2: A COL applicant that references the NuScale Power Plant design certification will develop a 20.4 supporting emergency response organization structure with defined roles and responsibilities to implement the strategies of the emergency operating procedures, severe accident mitigation guidelines, diverse and flexible coping strategies support guidelines (FSGs), and extensive damage mitigation guidelines. COL Item 20.4-3: A COL applicant that references the NuScale Power Plant design certification will develop and 20.4 describe at least one onsite and one offsite communications system capable of remaining functional during an extended loss of alternating current power including the effects of the loss of the local communications infrastructure. COL Item 20.4-4: A COL applicant that references the NuScale Power Plant design certification will develop, 20.4 implement, and maintain the training and qualification of personnel that perform activities in accordance with diverse and flexible coping strategies support guidelines (FSGs), severe accident mitigation guidelines, and extensive damage mitigation guidelines. The training and qualification on these activities will be developed using the systems approach to training as defined in 10 CFR 55.4 except for elements already covered under other NRC regulations.
Tier 2 1.8-20 Revision 2 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 20.4-5: A COL applicant that references the NuScale Power Plant design certification will develop drills 20.4 or exercises that demonstrate the ability to transition to one or more of the strategies and guidelines of the emergency operating procedures, diverse and flexible coping strategies support guidelines (FSGs), extensive damage mitigation guidelines, and severe accident mitigation guidelines using only the station communication equipment designed to be available following an extended loss of alternating current including effects of the loss of the local communications infrastructure. COL Item 20.4-6: A COL applicant that references the NuScale Power Plant design certification will develop and 20.4 describe the means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials to the environment including releases from reactor core and spent fuel pool sources.
Tier 2 1.8-21 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria
1.9 Conformance with Regulatory Criteria
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"
COL 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.
1.9.1 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
Tier 2 1.9-1 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria
1.9.2 Conformance with Standard Review Plan
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.
1.9.3 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.
1.9.4 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
Tier 2 1.9-2 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria
sections as they are updated. Operational experience from NRC Bulletins and Generic Letters not incorporated into the most recent applicable SRP six months before the 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. NuScale’s 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.
1.9.5 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.
1.9.6 References
1.9-1 NuScale Power, LLC, NP-RT-0612-023, Rev 1, Gap Analysis Summary Report.
Tier 2 1.9-3 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria
Table 1.9-1: Conformance Status Legend Conformance Status Description Code Conforms 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. Partially 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 Not 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. Departure 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 1Note 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.
Tier 2 1.9-4 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.7 8.3 12.3 Not Applicable Not Applicable Not Applicable Not Applicable 0.3 to be used for new 0.3 to be used theNPM containment. Not Applicable existing reactors; RG 1.183 is existing reactors; RG 1.183 of the COL applicant or licensee. In addition, seismic detectors cannot be installedthe contain- inside they are installed in 3.7.3 indicates so Section ment the RXB. safety-related emergency diesel generators. ties are the responsibilityCOL applicant. of the combustible gas control systems. tain1E distribution systems. The EDSS design Class guidance for independence of to the conforms standby power sources and their distribution sys- tems. specified15. in SRP Section reactors. - Applicable Not - This guidance is only applicable to BWRs. Partially Conformspowerelectrical Theonsite systems do not AC con- Not Applicable 12 Applicable Not lines No penetrate Partially Conforms Selection of specific equipment is the responsibility 4 Applicable Not TheNuScale design does not require or include 3 Applicable Not operational Site-specificactivi- programmatic and 3 Applicable Not Thecontainmentnot rely on vessel integrity does 22 Applicable Not This guidance is only applicable to BWRs. Applicable Not This RG pertains to Not Applicable Table 1.9-2: Conformance with Regulatory Guides with Regulatory Conformance Table 1.9-2: Primary Reactor ContainmentPrimary Reactor for tion Earthquakes Related Diesel Generators in Related Diesel Nuclear Power Plants sonnel for Nuclear Power Plants sonnel for centrations in Containment the Potential Radiological Conse- quences of a Loss of Coolant Acci- Boiling Water Reactors dent for the Potential Radiological Conse- quences of a Loss of Coolant Acci- Pressurized Water dent for Reactors the Potential for Evaluating Used Radiological Consequences of a for Line Break Accident Steam Reactors Boiling Water Between Redundant Standby (Onsite) Power Sources and Sys- Between Their Distribution tems RG Division Title Rev. Status Conformance Comments Section 1.111.12 Instrument Lines Penetrating the Nuclear Power Plant Instrumenta- 1.9 Application and Testing of Safety- 1.8Training of Per- and Qualification 1.7 Control of Combustible Gas Con- 1.31.4 for Evaluating Used Assumptions for Evaluating Used Assumptions 1.5 Assumptions 5 - Guide Safety 1.6Independence Guide 6 - Safety
Tier 2 1.9-5 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.9 5.4 3.2 9.1 9.2 14.2 11.5 3.5.2 Not Applicable Not Applicable olant pumps. such protection is not ents, but as structures, for reactor for reactor co ototype reactors are appli- th the capability to vent capability th the M is classified as a prototype are theresponsibilityof the COL applicant or licensee. cant or licensee. rely on reactor coolant pumps.The NuScale design uses passive naturalthe primary cool- circulation of ant, eliminating the need in accordance with RGThus, the portions 1.20. of thisapply RG which to pr cable to the first operationalthe first NPM. After NPM is qualified as a valid prototype, subsequent NPMs are classifiednon-prototype ascategory I and non-prototypetheportions apply. The of the RG from RG of departures identification is the 1.20 responsibilitylicensee.or of the COL applicant complies with Regulatory Position C.8, Makeup Position complies with Regulatory the withininventory of water large Water by the ultimate forming the Seismic Category I structures Group C, by the separate Quality heat sink (UHS) and Posi- line. For Regulatory I makeup Seismic Category tionCooling, C.9, PoolstructuresUHS pool the con- taining the inventorymakeup of water credited for spentcooling fuel during accident conditionsmeet Seismic Category I requirem they are not designedQuality to Group C require- the reactor buildingaddition, ventilationments. In is not credited wi system protect steam or moisture to theatmosphere safety-related components from high temperatures moisture levels because and design. required for the 3 Conforms The first operational NP 2 Partially Conformsoperational Site-specific,aspects programmatic and CnomNn.7.2 0ConformsNone. 1 Applicable Not This guidance is theresponsibility of the COL appli- 1 Not Applicable This guidance is applicable only to PWR designs that 2 Partially Conformsdesign of the The new and spent fuel storage facility Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table ment Program for Reactor Inter- ment Program for nals During Preoperational and Testing Initial Startup Reporting Radioactive Material in Reporting Radioactive and Effluents Gaseous and Liquid Solid Waste System Actuation Functions Actuation System Nuclear Power Plant for grams Integrity Design Basis RG Division Title Rev. Status Conformance Comments Section 1.20 Vibration Comprehensive Assess- 1.21 Measuring, Evaluating, and 1.221.23 Periodic Testing of Protection Meteorological MonitoringPro- 1.14 Flywheel Reactor Coolant Pump 1.13Facility Fuel Storage Spent
Tier 2 1.9-6 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.2 5.2 5.4 6.2 9.1 9.2 9.3 10.3 10.4 Not Applicable Not Applicable Not Applicable ing system to transfer heat to the ultimate heat sink. heat the ultimate heat to transfer to system ing TheNuScale design uses a passive cooling system. cable to a specific component is described through- out the FSAR. licensees of existing reactors thatlicensees are not authorized of CFR 10 under use the alternative source term to RG to be used in lieu of specified is 1.183 RG 50.67. reactorsexisting autho- (and for new reactors 1.25 torized use the alternative sourceunder10 term CFR 50.67). COL applicant or licensee. 3 Applicable Not RG does not applyuse a passiveplants to cool- that 4 Conformsquality The group classificationappli- from RG 1.26 0 Applicable Not by source terms used This RG pertains to TID14844 0 Applicable Not Site-specific guidance is the responsibilityof the Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Power Plants Standards for Water-, Steam-, and Standards Radioactive-Waste-Containing ComponentsPower of Nuclear Plants the Potential Radiological Conse- Handling Acci- of a Fuel quences dent in the Fuel Handlingand Boiling and Facility for Storage Pressurized Water Reactors the Potential Radiological Conse- quences of a Pressurized Water Reactor Gas Storage Radioactive Tank Failure RG Division Title Rev. Status Conformance Comments Section 1.27 Ultimate Heat Sink for Nuclear 1.26 QualityClassifications Group and 1.25 for Evaluating Used Assumptions 1.24 for Evaluating Used Assumptions
Tier 2 1.9-7 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.2 8.3 5.2 6.1 4.5 5.2 5.3 5.4 6.1 7.0 7.2 3.2 14.3 17.1 17.5 Not Applicable 1-2008 and and NQA-1a-2009 icableoffsite to theand onsite AC IEEE Std. 336-1971 for the installa- for the 336-1971 Std. IEEE bed in Staff Regulatory Guidance Guidance Regulatory bed in Staff power systems. The EDSS conforms to RG 1.32 to the toto RG 1.32 conforms EDSS systems. The power describedextent in Section 8.3. tion, inspection,and testing of instrumentation and design is based onelectric equipment. The NuScale addenda, as NQA-1a-2009 and the NQA-1-2008 and 1.28, Rev. 4. NQA-1-2008 RG in endorsed NQA-1a-2009 (Subpart 2.4) reference IEEE Std. 336- The sub- 336-1971). Std. to IEEE (as opposed 1985 stantive content andcontained intent of RG 1.30 is NQA- of in Subpart 2.4 IEEE Std. 336-1985, which is applicable to the NuS- applicable to the is which 336-1985, Std. IEEE and NQA-1a-2009. design per NQA-2008 cale C.1.a through C.1.h is designated as Seismic Cate- designated is C.1.a through C.1.h C.1.i Guidance Regulatory SSC that meet Staff gory I. are designated Seismic Category II rather than Seis- mic Category I. The seismic classification from RG applicable1.29 to a specific component is described the throughout FSAR. - Applicable Not This RG endorses 3 Partially Conformsappl RG 1.32 is not CnomNn.4.5 4ConformsNone. 4 Conforms None.5 Conforms Partially descri Each SSC 3.13 Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Nuclear Power Plants Stainless Steel Weld Metal ance RequirementsInstal- for the lation, Inspection,and Testing of Instrumentation and Electric Equipment ria (Design and Construction) Nuclear Power Plants RG Division Title Rev. Status Conformance Comments Section 1.32 for for Power Systems Criteria 1.31in Content of Ferrite Control 1.30 Guide 30 - Safety Quality Assur- 1.28Crite- Quality Assurance Program 1.29 Seismic Design Classification for
Tier 2 1.9-8 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 4.5 5.2 5.3 6.1 8.3 5.3 6.1 14.2 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable able except for its specification of able except for guide are applicable preopera- to applyingflushingcleaning1.37 for and of fin- RG ished surfaces. has been withdrawn by the RG 1.37 NRC. Class 1E motors. Class 1E tional testing of divisional EDSS load groups. It does not apply to NuScale AC power systems or the EDNS. mal insulation on RCPB or CNV components. (i.e., does not use concrete in its design). (i.e., (i.e., does not use concrete in its design). (i.e., cant or licensee. - Partially Conforms Portions of this - Applicable Not nonmetallic TheNuScalether- design does not use - Applicable Not containment TheNuScale design uses a steel vessel CnomNn.5.2 1ConformsNone. 1 Partially Conforms This RG is applic 1 Applicable Not continuous TheNuScale design does not use duty CnomNn.5.3 1ConformsNone. 3 Applicable Not containment TheNuScale design uses a steel vessel 3 Applicable Not This guidance is theresponsibility of the COL appli- Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table dant Onsite Electric Power Sys- tems to Verify Proper Load Group Assignments Cladding of Low-Alloy Steel Com- Steel Low-Alloy of Cladding ponents of Stainless Steel Safety-Related Motors for Nuclear Safety-Related Power Plants for Austeniticfor Stainless Steel for Inspection of Prestressed Con- crete Containments erties TendonsUngrouted in Pre- stressed Concrete Containments Requirements (Operation) RG Division Title Rev. Status Conformance Comments Section 1.41 Testing of Redun- Preoperational 1.431.44 Control of Stainless Steel Weld Controlof the Processing and Use 1.40 Qualification of Continuous Duty 1.36 Nonmetallic Thermal Insulation 1.35.1 Prestressing Determining Forces 1.341.35 Control of ElectroslagProp- Weld of Inspection Inservice 1.33 Quality Assurance Program
Tier 2 1.9-9 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.6 5.2 6.2 9.3 6.1 11.5 14.2 14.3 lity to identify or quan- g transport delay time) is is treated as unidentified as unidentified is treated systems to detectsystems to leakagethe containment: into leakage col-containment pressure monitoring and lection. Both leakage detectionReg- satisfy methods a) C.2.2 in RG 1.45: C.2.1 and ulatory Positions reactor containmentto the primary from leakage unidentified sourcescan be detected, monitored, gpm; and; b) for rates ≥ 0.05 quantified and response time (not includin for a leakageless than rate greater than one hour PositionC.2.4 is satisfiedone gpm. Regulatory because the containment pressure method is capa- of performing its function following a seismicble shutdown (i.e., require plant event that does not is satisfiedvacuum pump remains functional). C.2.5 because both methods permit calibration and test- radiation detec- plant operation. Finally, ing during ancondenser early torsline provide in the CES vent leakage consistent with Regulatory indication of RCS Position C.2.3. All leakage limited capabi because of the tify RCS leakage. CnomNn.7.2 5.2 1ConformsNone. 1ConformsNone. 1 Conforms1.45 guidancedesign The by using two satisfies RG Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Indication for Nuclear Power Indication for Nuclear Safety Systems Plant for Welding of Low-Alloy Steel Responding to Reactor Coolant System Leakage RG Division Title Rev. Status Conformance Comments Section 1.471.50 Bypassed and Inoperable Status Control of Preheat Temperature 1.45 and on Monitoring Guidance
Tier 2 1.9-10 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.3 15.1 15.2 15.5 11.2 Not Applicable ions. However, these sys- operational aspects (e.g., maintenance of safety- operational aspects that are the related coatings) responsibility of the COL applicant or licensee. (ESF) filter and atmosphere cleanup(ESF) filter and systems product removaldesigned for fission in a post- design basis accident environment. The NuScale atmosphere filter and rely on ESF not does design cleanupof a systems to mitigate the consequences design basis accident. Nonsafety-related normal ventilation systems provide atmosphere cleanup capability,design, as test- meets the necessary, that These 1.140. ing, and maintenance guidelines in RG containment,systems provide appropriate confine- ment, and filtering to limit releases of airborne radio- the environmentandactivity to during normal postulated accident condit a at accident an following required not are tems NuScale Power Plant, and accordingly receive no the radiological con- of determination credit in the sequences of an accident. CnomNn.7.1 2ConformsNone. 2 Partially Conforms Applicable except for portions of this RG that govern 4 Applicable Not This guidancefeature addresses engineered safety Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Criterion to Safety Systems to Safety Criterion to Nuclear Applied Coatings Power Plants Criteria for Air Filtration and for Air Criteria AdsorptionUnits of Post-Acci- dent Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants RG Division Title Rev. Status Conformance Comments Section 1.53 Application of the Single-Failure 1.54 Service Level I, II, and III Protective 1.52 Inspection, and Testing Design,
Tier 2 1.9-11 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.7 3.8 8.1 8.3 7.2 3.8 6.2 3.12 3.11 14.3 3.8.2 Not Applicable Not Applicable m flood height (includ- more applicable to the load combinations speci- typical RPV load combi- load RPV typical e 1997 version, including version, e 1997 the design envelops the RG e RG guidance that endorses 1.63 IEEE-317-1983 is applicable. IEEEis applicable. IEEE-317-1983 is used 741-1997 penetra- of electrical protection circuit for external tion instead of IEEE assemblies as 741-1986 endorsed by RG 1.63. Th additionalthe design enhancements, is consistent with RG 1.63. native damping value for the NPM substructure was NPM substructurevaluenative damping for the determined. (CSDRS) was not developed using RG 1.60. However, was (CSDRS) it is demonstrated that anchored to 0.1g. spectra 1.60 above the probableabove maximu ing wind-inducedrun-up). wave 10 CFR since per 10 50.34(f)(3)(v), CFR and 50.34(f) 10 CFRcertification a design 52.47(a)(8), applicant havedoes to show compliance with not 10 CFR of Use 50.34(f)(3)(v). vesselsfor Classnations 1 is CNV than using the load combinationsin specified RGbecause of the increased 1.57 quality of the fabri- cation, inspection,required by and testing ASME for a Class 1 NB Subsection III, Code, Section vessel. Additionally,use of load the combinations for vesselsClass 1 is more applicableto the contain- ment vessel than using the the quality of because of fied in RG 1.57 testing required by fabrication, inspection, and for a Class 1 NB Subsection III, Code, Section ASME vessel. CnomNn.7.1 1ConformsNone. 3 Partially Conforms Theportion of th 1 Conformsan In accordance with theguidanceof alter- RG 1.61, 2 Applicable Not TheCertified Seismic Design Response Spectra 2 Applicable Not TheNuScale design assumes the NPP is located 2 Partially Conforms Applicablefor reference to except Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Actions Containment Structures for Nuclear Power Plants Designof Nuclear Power Plants mic Design of Nuclear Power Plants Power Plants binations for Primary Metal Reac- tor Containment System Components RG Division Title Rev. Status Conformance Comments Section 1.621.63 Manual Initiation of Protective Electric Penetration Assemblies in 1.61Values for Seismic Damping 1.60Response Spectra for Seis- Design 1.59 Basis Floods Design for Nuclear 1.57 Limits Design and Loading Com-
Tier 2 1.9-12 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.8 4.4 5.4 8.2 8.3 5.3 12.3 14.2 10.4 14.2 14.3 14.2 14.2 14.2 3.13 9.3.2 Not Applicable surevessel (RPV) bolting e design. Site-specifice design. pro- ns addressed by RG 1.65 e Design Specific Review Stan- dards (DSRS) are used. dards (DSRS) BWR-specific or address specific PWR design fea- specific PWR BWR-specific or address tures not in theNuScale design. aspectsincluding test performance, test report preparation, and records retention, which are the responsibilitylicensee.or of the COL applicant aspects,including test performance and records the COL of responsibility which are the retention, applicant or licensee. radiation shields.this guid- Site-specific aspects of andance, including development implementation responsi-of a radiationtest program, are theshield applicantofbility the COL or licensee. are BWR-specific or address specific design PWR SSC NuScal not in the features activitiesgram implementation are the responsibil- licenseeor ity of the COL applicant that references certified design. NuScale the applicant. The reactor pres is material The material is described in Table 5.2-4. the concer to not subject Positions 1(a)(i) and 2(b). Therefore, these Positions material. bolting to the RPV apply not do 3 Applicable Not and NuScal RG 1.206 22 Partially Conforms Thisare RG is applicableaspects that except for 1 Partially Conforms This guidance is applicablesite-specific except for 1 Partially Conforms This guidance is applicablesite-specific except for Partially Conforms This guidance is applicable to the design of concrete 4 Partially Conforms This guidance is applicableaspects except for that 1 Partially Conforms Inservice inspection is the responsibilityof the COL Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Safety Analysis Reports for Analysis Reports Safety Edi- (LWR Power Plants Nuclear tion) sate and Feedwater Systems for sate and Feedwater Reactors Light-Water Shutdown Remote Demonstrate CapabilityWater-Cooled for Nuclear Power Plants ment and Control AirSystems Shield Testing for Nuclear Generic Power Plants Cooled Nuclear Power Plants Nuclear Cooled Reactor Vessel Closure Studs Vessel Closure Reactor RG Division Title Rev. Status Conformance Comments Section 1.70Format and Content of Standard 1.68.1 of Conden- 1.68.2 Test Program Initial Test Program Initial Startup to 1.68.3 Testing of Instru- Preoperational 1.69Radiation Shields and Concrete 1.68Water- for Test Programs Initial 1.65 Materials and Inspections for
Tier 2 1.9-13 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 2.3 3.3 3.5 3.8 5.0 7.2 8.3 9.5 4.5 5.2 6.1 3.11 14.3 Not Applicable cept for portionscept that llation, and operation of a ication of standards for ication idance is applicable ex apply to high temperature gas-cooled reactor designs. is parameters. The COL applicant or licensee design confirming the characteristics. responsible for pond applicationsthe UHS design). (or for aspects, including specif weld fabrication and repair that are performed during construction,insta of the responsibility nuclear facility,which are the COL applicant or licensee. CnomNn.7.1 13ConformsNone. Partially Conforms Thegu 1 Conformscharacteristics are used as1 (bounding) Region 2 Applicable Not use fiberglass Thedesign does piping in spray not 1 Partially Conforms This guidance is applicablesite-specific except for Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Related Actuators in Nuclear Power Plants Systems trical Safety Power Missiles for Nuclear nado Plants Fiberglass-Reinforced Thermoset- ting Resin Limited Accessibility RG Division Title Rev. Status Conformance Comments Section 1.731.75for Safety- Qualification Tests Elec- of for Independence Criteria 1.76 Design-Basis Tornado and Tor- 1.72 from Made Pond Piping Spray 1.71 Qualification for Areas of Welder
Tier 2 1.9-14 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.3 3.2 6.4 9.4 14.2 15.4 15.0.3 B. The radiological cri- B. The . Per SRP Section 15.0.3, SRP Section . Per ge refers to SSC design design refers to SSC ge identified in SRP Section this guidance require RG as out of date and in date of as out RG .g., amount and location and cladding failureand cladding crite- Section 15.0.3 supersedes Section 15.0.3 e criteria provided SRP e criteria provided RG is applicableis to theNuScale RG design, but the literal langua literal the but design, features not in the NuScale design. For example, the pres- use high pressure or low not design does ECCS as described in this pumps sure safety injection provides core design the ECCS guidance. Rather, decay heat removal by steam condensation and nat- ural reactor coolant recirculation.Nevertheless, pre- operational testingbeperformed will on the ECCS in this guidance. satisfies the intent of a manner that RG prescribesMuch of this preoperational test implementation activities that are the responsibility of the COL applicant. logical consequence analysis undersubhead-Section I, Areas of Review, Item 10 design certi- reviewInterfaces, for the of ing Review fication applications, SRP radiologicalthe analyses, assumptions, acceptance criteria, and methodologies in RG 1.77. guidance which references 15.4.8, The NRC has identified this of revision. The fuel need ria are superseded by th Appendix 4.2, (NUREG-0800) teria1.183. How- are superseded by the criteria in RG generalever, the approach still and intent of RG 1.77 apply and are used in Section 15.4.8 analyses. ity design within the scope of the NuScale design are applicable. Otheraspects of (e information site-specific and room, the control of toxic chemicals relative to spec- or factors) dispersion atmospheric site-specific ify operational,planning programmatic emergency aspectsof are the responsibility the activities. These COL applicant or licensee. - Conforms Partially for radio- assumptions to RG pertain of this Portions 2 Conforms Partially this intent of The 1 Partially Conforms Aspects of this habitabil- RG related to control room Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table gencySystems for Core Cooling Pressurized Water Reactors Nuclear Power Plant Control Haz- During a Postulated Room ardous Chemical Release a Control Rod Ejection Accident Ejection Accident Rod a Control Reactors Pressurized Water for RG Division Title Rev. Status Conformance Comments Section 1.79 Testing of Emer- Preoperational 1.78 EvaluatingHabitability the of a 1.77 for Evaluating Used Assumptions
Tier 2 1.9-15 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.3 Not Applicable with the exception that nt to the AC power systems. As systems. AC power to the nt downstream clogging). reaction productsdownstream). cleanness programs, monitoring/sampling term, debris,insulationfor latent selection, restriction on steel).coatings and cladding of carbon 1.82 regulatory positions that addressdesign1.82 regulatory positions that the criteria, performance standards, and analysis long-term for sources water to related methods cooling. However, the NuScale design differs from addresses. guidance the designs system the TheNuScalewith the guidance design complies debris generation, transport, with respect to debris, downstream, and coating debris, latent chemical effects. The NuScale design is passive and includedoes pumps, sumps, suction strainers, not and the design debris interceptor, or trash racks, negates the potential effect of non- or minimizes condensables on coolant flow to the core. The require operatordesign doesaction not NuScale to mitigate debris accumulation. NuScaleThe design does not comply with C1.1 regulatory position with the intent of the does comply NuScale following regulatory positions: •the possibility Position C1.1.1.9 (assessment of of •debris and chemical (buildup of Position C1.1.1.10 • (minimization of debris source Position C.1.1.2 described in Section 8.3, the EDSS conforms to por- conforms Section 8.3, the EDSS in described tions of RG 1.81. - Applicable Not applicable RG 1.79.1 is to BWRs only. Not Applicable 4 Conforms Partially RG of intent the with complies design NuScale The 1 Applicable Not not releva is RG 1.81 Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Recirculation Cooling Following a Loss-of-Coolant Accident down Electric Systems for Multi- for Systems Electric down Unit Nuclear Power Plants Core Cooling Systemsfor New Boiling-Water Reactors RG Division Title Rev. Status Conformance Comments Section 1.82 Long-Term Sources for Water 1.81 Shared Emergency and Shut- 1.79.1 Emergency of Test Program Initial
Tier 2 1.9-16 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 4.5 5.2 3.13 Not Applicable theexception that the (prototypical head loss (upstream effects) or effects) (upstream applicable to the design) to the design) applicable evaluation of long term with the NuScale design reactor operating licenses. Positions C1.1.3 and C1.1.4 are not applicable and C1.1.4 C1.1.3 Positions not rely on because the NuScale design does of consequences the action to mitigate operator debris accumulation and does not include active devices or systems to prevent debris accumulation. NuScaleThe design does not comply with regulatory positionC.1.2 (alternative water sources for inoperable strainers). TheNuScalewith the intent design complies of C.1.3 ( regulatory position as recirculation capability the following: with the exception of • (net positive suctionhead) Position C.1.3.1 •consistentthat are not Portions of position C.1.3.2 NuScaleThe design does not comply with regulatory positions C1.3.7 C.1.3.9 (strainer structural integrity). NuScaleThe design does not comply with C1.3.12 regulatory position testing). NuScaleThe design does not comply with C.2 with regulatory position intent of chemical reaction effects (position 2.2) is met. NuScaleThe design does not comply with C.3. regulatory position - Applicable Not Thisnuclear RG governs terminating the process for 36 Conforms None. 3.12 Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Licenses for Nuclear Reactors Licenses for Nuclear Code Case Acceptability, ASME Code III Section RG Division Title Rev. Status Conformance Comments Section 1.86 Termination of Operating 1.84 Fabrication, and Materials Design,
Tier 2 1.9-17 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.8 3.9 3.10 3.12 3.11 3.8.2 Appendix 3C Appendix Not Applicable Not Applicable Not Applicable Not Applicable em, etc.); and (2) refer- em, etc.); and gas-cooled reactors, liq- fied as an applicable RG in DSRS to elevated-temperature reactors cific evaluations and is the responsibility of the COL cific evaluations andresponsibility is the applicant or licensee. Section 8.1. BWR-specific or related to SSC that are notare relevant that BWR-specific or related to SSC design (e.g.,contain-the NuScale ice condenserto ment, containment spray syst term 1.4 for source term, since the ence to RG for RG 1.183 1.4 are superseded by RG of provisions new reactors. rate a pre-stressed concrete containment structure containment NuScale The tendons. with grouted not use concrete or grouteddoes vessel is steel (i.e., tendonsdesign). in its such as high-temperature uid-metal fast-breederand gas-cooled fast- reactors, breeder reactors. CnomNn.3.7 3ConformsNone. 11 Applicable Not This RG is not identi Applicable Not This RG is applicable only to BWR designs. Not Applicable 2 Applicable Not This guidanceperformance governs the of site-spe- 1 Partially Conforms This RG is applicable(1) aspects that except for: are 2 Applicable Not designs incorpo- that LWR is applicable to This RG 1 Applicable Not This RG applies Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Spatial Components in Seismic Response Analysis Sources Valve Leakage Control Systems Boiling Water Reactor Nuclear for Power Plants lated to Occur on Transportation Transportation on to Occur lated Power Plants Near Nuclear Routes Certain Electric Certain Equipment for Nuclear to Safety Important Power Plants stressed Concrete Containment Tendons with Grouted Structures Class 1 Components in Elevated- Class Temperature Reactors (Supple- ment to ASME Section III Code 1594, 1595, and Cases 1592, 1593, 1596) RG Division Title Rev. Status Conformance Comments Section 1.92Responses and Modal Combining 1.931.96 Power Electric of Availability Main Steam Isolation of Design 1.91 EvaluationsExplosions of Postu- 1.89 Environmental Qualification of 1.90 Inservice Inspectionof Pre- 1.87Construction of for Guidance
Tier 2 1.9-18 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.9 5.2 5.4 7.1 7.2 8.3 3.10 14.3 11.5 12.3 14.3 3.11 Appendix 3C Appendix Appendix 3C Appendix when site-specific spectra exceed the certified and (2) quali- spectra (e.g., Position C1.2.1.g); design equipmentnew in older and replacement fication of Position plants (e.g., A46 issue safety unresolved Not applicableC.1.2.2.j). to electrical equipment. Site-specific guidance is the responsibilityof the COL applicant. RG 1.100 endorses ASME QME-1 complies with the non- NuScale 2007. the followingmandatory Appendix QR-B with exceptions: Qualifi- of Identification and Specification QR-B5200, cation Requirements,(g) material activation energy. Selection of Qualification Methods for QR-B5300 lifeof determination and recording of shelf nonmetallics. Documentation,QR-B5500 (h) shelf life preservation requirements. describes the exceptionsAppendix 3C cited. ments in Section 6.6 of IEEE Std 497-2002 for Type B Type for Std 497-2002 6.6 of IEEE ments in Section rather power reliable highly with C variables and than with Class 1E. CnomNn.5.3 0 Applicable Not 2ConformsNone. This RG is applicable only to BWR designs.3 Not Applicable Partially Conforms This RG is applicable except for aspects related to: (1) 4 Conforms Partially require- supply power satisfies design NuScale The Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table the Potential Radiological Conse- quences of a Radioactive Offgas Boilinga Water Failure in System Reactor tor Vessel Materials Mechanical Equip- and Active ment and Functional Qualifica- tion of Active Mechanical Equipment for Nuclear Power Plants InstrumentationNuclear for Power Plants RG Division Title Rev. Status Conformance Comments Section 1.98 for Evaluating Used Assumptions 1.991.100 Radiation Embrittlement of Reac- Seismic Qualification of Electrical 1.97 Criteria for Accident Monitoring
Tier 2 1.9-19 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.2 2.4 3.4 3.8 11.2 11.3 15.1 15.2 15.4 15.5 15.6 Not Applicable Not Applicable Not Applicable ight (including wind es updated guidance for guidance updated es an allowable value to le except for specification of site- le except for le only to LWR designs that use a es use the safety-related setpoints safety-related es use the prestressed concrete containmentThe structure. containment vessel is a steel (i.e., does not use concrete or pre-stressed tendons in its design). specific information (e.g., meteorological data). Site- specificinformationthe responsibility of COL is applicant. protection devices to ensure that safety-related protection devicesensure that to valvesmotor-operated perform their safety func- tion. The NuScale design does not use safety-related motor-operated valves. described in Chapter 7. This RG endorses ISA- in described however, the NuScale Instrument 67.04.01-1994, Methodology Technical Report (TR-0616- Setpoint applies the guidance contained49121) in ISA- ver- that the 1994 is A key difference 67.04.01-2006. uses sion of ISA-67.04.01 determine instrument channel operability during surveillance testing and calibration. ver- The 2006 provid sion of ISA-67.04.01 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. probable maximum flood he run-up). wave induced proposingsite to a power plant that meets thedefi- nition of co-located. 1 Partially Conforms This RG is applicab 2 Applicable Not This RG is applicab 2 Not Applicable This RG governs the application of thermal overload 3 Partially Conforms15 analys Chapter 1 Applicabledesignis The assumes located the NPP above the 5 Applicable Not This RG is theresponsibilityof the COL applicant Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Man from Routine Releasesof Purpose for the Effluents Reactor of Evaluating Compliance with 10 CFR Part 50,Appendix I ing for Prestressing Tendons in Tendons for Prestressing ing Containment Structures Electric Motors on Motor-Oper- Electric ated Valves Instrumentation Power Plants and Preparedness for Nuclear and Preparedness Power Reactors RG Division Title Rev. Status Conformance Comments Section 1.109 Calculation of Annual Doses to 1.107 Grout- Qualification for Cement 1.106Overload Protection Thermal for 1.105 Setpoints for Safety-Related 1.102 Flood Protection for Nuclear 1.101 Emergency ResponsePlanning
Tier 2 1.9-20 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.5 9.1 2.3 3.3 2.3 18.5 11.2 11.3 11.2 11.3 Not Applicable n B.1, the ability of the of the ability B.1, n guidance that implement le except for specification of site- le except for specification of site- le except for COL applicantlicensee.orCOL Consistent with the dis- cussion in RG 1.114, Sectio COL applicant to meet this guidance is facilitated by controlthe room design(including and layout the designatedPositionin surveillance area described Portions of this C.1.3). CFR requirements of 10 operator staffing are not and (iii) (e.g., Position C.1.5) 50.54(m)(2)(i) applicable to COL applicants. extreme wind loads will not adversely affect the loads will extreme wind Reactor Building or the Seismic Category I portion of Controltheof Building is the responsibility the COL applicant or licensee. performance of a site-specificperformancecost-benefit analysis. Site-specific information is the responsibilityof the COL applicant or licensee. dispersion of radioactivedispersion liquid effluents from com- of ponent failures, in accordancewith BTP 11-6. provides an approvedBecause the NuScale facility design mitigative feature (metal-lined concrete dike around the PSCS storage tank), such an analysis is not required. specificdispersion data. Site-specific information is theresponsibilityof the COL applicant or licensee. specific information (e.g., meteorological data). Site- specificinformationthe responsibility of COL is applicant or licensee. 3 Partially Conforms Site-specific guidance is the responsibilityof the CnomNn.3.5.3 2ConformsNone. 2 Conforms to exposed structures nearby Confirmation that 11 Partially Conforms This RG is applicableaspects related except for to Partially Conforms1 This RG is applicab Partially Conforms This RG is applicab 1 Applicable Not Thisaquatic RG provides guidance for analyzing the Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Controls and toSeniorOperators Room of a Nuclearin the Control Power Unit siles Eventsfor Nuclear and Missiles Power Plants waste Systems for Light-Water- Nuclear Power Reactors Cooled spheric Transport and Dispersion Routine in Effluents Gaseous of from Releases Light-Water- Cooled Reactors and in Gaseous Materials active from Light- Effluents Liquid Water-Cooled Nuclear Power Reactors Effluents from Accidental and Accidental from Effluents the Releases for Reactor Routine of Purpose Implementing Appen- dix I RG Division Title Rev. Status Conformance Comments Section 1.114the Operators at to Guidance 1.1151.117Turbine Mis- Against Protection ProtectionExtreme Wind Against 1.110Rad- Analysis for Cost-Benefit 1.111 Methods Estimating Atmo- for 1.112 Radio-of of Releases Calculation 1.113 Dispersion of Aquatic Estimating
Tier 2 1.9-21 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.3 8.3 3.7 7.2 8.3 14.2 3.12 14.2 Not Applicable Not Applicable ciated inservice inspec- does not require hydraulic struc- VRLA batteries. NuScale VRLA batteries. applies uses VRLA batteries; thus batteries; uses VRLA IEEE Std amendment. with the 2014 1188-2005 IEEE Std is applied. 1187-2013 vice inspection and surveillance program for dams, slopes, canals,and other water-control structures associated withcooling water systems emergency or flood protectionnuclear of power plants. Water asso and structures control tion and surveillance programs are site-specific details. Site-specific guidanceresponsibility is theof COLthe applicant or licensee. COL applicant or licensee. tures. COL applicant or licensee. 3 Partially ConformsEDSS uses The 2 Partially Conforms TheEDSS CnomNn.3.9 4.2 2ConformsNone. 22ConformsNone. Not Applicable The NuScale design 1 Applicable Not This guidanceof an inser- development governs the 1 Partially Conforms Site-specific guidance is the responsibilityof the 3 Partially Conforms Site-specific guidance is the responsibilityof the August 1976 Conforms None. 5.4 Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table ear Power Plants Replacement of Vented Lead- Replacement of Acid Storage Batteries for Nuclear Power Plants tion oftion Vented Lead-AcidStorage Batteries for Nucl binations for Class 1 Linear-Type for Class 1 binations Supports Operation of HydraulicStruc- Systems for Nuclear tures and Power Plants for the Analy- Methods Statistical Densificationsis of Fuel Structures Associated with Nuclear Power Plants PWR Steam Generator Tubes PWR Spectra for Seismic Response Equip- Floor-Supported of Design ment or Components and Protection Systems RG Division Title Rev. Status Conformance Comments Section 1.129 Maintenance, Testing, and 1.128 InstallationDesign and Installa- 1.1241.125Loading Com- Service Limits and for Design and Models Physical 1.1261.127 An Acceptable Model and Related Inspection of Water-Control 1.1211.122 Degraded for Plugging Bases Development of Floor Design 1.118 Periodic Testing of Electric Power
Tier 2 1.9-22 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.2 3.2 9.4 11.3 12.3 14.2 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable nstruction, testing, and and testing, nstruction, RGadherence to ANS through 1.141 N271-1976. 56.2-1984, Note: the provisions of ANSI/ANS Sectionto penetrations with 3.6.5 are applied both CIVs outside containment that serve non-ESF process systems. ble. Aspects related co to related Aspects ble. repairs are the responsibility of the COL applicant or licensee. investigation activities. Site-specific guidance is the responsibilitylicensee.or of the COL applicant emergency diesel generators. emergency diesel use concrete containments. The NuScale design uses a steel containment vessel. activities. Compliance with site-specificactivities. Compliance with guidance is theresponsibilityof the COL applicant or licensee. lation flow combineddesigna with that only has 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 SectionFSAR. of the 4.4 part of site selection.part of Site-specific guidance is the responsibilitylicensee.or of the COL applicant 1 Conformsrequirements of to the NuScale design conforms The 2 Partially Conformsaspects Design-related of this guidance are applica- 2 Applicable Not Thisto site-specific guidance is related laboratory 2 Not Applicable The design does not rely on or include safety-related 3 Not Applicable This guidance is applicable only to LWR designs that 3 Applicable Not program operational site-specific governs RG This 1 Applicable Not circu- natural from resulting velocities fluid low The CnomNn.3.9 3ConformsNone. 2 Applicable Not Thisas RG governs site investigations performed Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table s of Normal s for Standby Die- for Fluid Systems Criteria for Air Filtration and for Air Criteria Adsorption Unit Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants and Rocks for Engineering Analy- Rocks and sis and Designof Nuclear Power Plants sel Generators tions, Materials, Construction, Construction, Materials, tions, of Concrete Contain- and Testing ments Personnel at Nuclear Power Plants the Primary System of the Light- Water-Cooled Reactors binations for Class 1 Plate-and-for Classbinations 1 Shell-Type Supports tions of Nuclear Power Plants tions of RG Division Title Rev. Status Conformance Comments Section 1.141 Containment Isolation Provisions 1.140 Inspection, and Testing Design, 1.138 Laboratory Investigations of Soils 1.137 System Fuel-Oil 1.136Limits, Loading Combina- Design 1.134 Medical Evaluation of Licensed 1.133 Loose-Part Detection Program for 1.130Loading Com- Service Limits and 1.132 Site Investigations for Founda-
Tier 2 1.9-23 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.2 5.2 6.6 3.2 3.3 3.5 3.7 3.8 11.2 11.3 11.4 14.3 9.2.6 Not Applicable Not Applicable 7 with exceptions. The tions C.1 and C.14 related and C.14 tions C.1 RG regarding installation regarding RG es and conduct of licensed opera- guidance is applicable guidance is but the lan- inservice inspections per the ASME RG governs design and installation of safety- installation design governs and RG related instrument sensinginstrumentrelated lines in nuclear power this of aspects The plants. criteria are the responsibilityCOL applicant of the or licensee. tor training and qualification are the responsibility of COLthe applicant or licensee. effects. For the short distances that may be used for used for be may that For the short distances effects. to used is Guide 1.194 Regulatory and LPZ, EAB determine representative atmospheric dispersion factors. applicantBPVC is the responsibilitythe COL or of licensee. blowdown systems are not applicable to the design. Radioactive waste management system design crite- Construction, ria specified in this RG are applicable. installation, and testing criteria are the responsibility of the COL applicant or licensee. guage endorses ACI 349-199 guage endorses ACI standard has been used. the ACI 349 version of 2006 Posi Aspects of Regulatory concretetostructures within containment are not applicable to the design. The containment vessel is a steel component, and does not use interior concrete structures. 4 Applicable Not Simulation faciliti 1 Not Applicable This RG does not include modeling of building wake 2 Partially Conforms TheaspectsgeneratorRG related to steam of this 2 Conforms Partially this intent of The 17 Partially Conforms Performance of Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Acceptability,XI, ASME Section Division 1 Facilities for Use in Operator Training,Examinations, License and Applicant Experience Requirements for Potential Accident Conse- Potential Accident for quence Assessmentsat Nuclear Power Plants WasteSystems, Management Components and Structures, Installed in Light-Water-Cooled Nuclear Power Plants tures for Nuclear Power Plants (Other than Reactor Vessels and Containments) RG Division Title Rev. Status Conformance Comments Section 1.151 Instrument Sensing Lines 1 Partially Conforms This 1.1471.149 Inservice InspectionCode Case Nuclear Power Plant Simulation 1.145 Atmospheric Dispersion Models 1.143 Radioactive Guidance for Design 1.142Struc- Concrete Safety-Related
Tier 2 1.9-24 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.1 7.2 5.4 6.2 8.3 8.4 9.3 9.5 8.3 3.11 14.3 10.3 14.2 Not Applicable Not Applicable kes some tests not appli-some testskes not ations are notations used. Not Applicable ctiveness of maintenance activi- development lifecycle differs from conforms to the aspects of the RG asaspectsto the conforms it of the RG ties is theresponsibilityCOL applicant. of the the conceptual waterfall in RGlifecycle 1.152. The be will model lifecycle RG the from tasks applicable mappedthe I&C development to lifecycle. Compli- is condi- 7-4.3.2-2003 IEEE of ance with Clause 5.5 tioned by thechoicefield programmable of gate array technology, which ma cabletimer tests). (e.g., calculationtests, watchdog sibility of the COL applicant. pertains to passive plant designs. -- Applicable Not calcul Best estimate ConformsDC system batteries The are non-Class1E. 3.11 3 Applicable Not Monitoring the effe 3 Partially Conforms TheNuScale I&C 1 Conforms None.2 Applicable Not activities are the respon- Decommissioning funding 3.11 Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Maintenance atMaintenance Nuclear Power Plants Safety Systems of Nuclear Power Plants AssembliesNuclear for Power Plants System Core Cooling Emergency Performance for Decommissioning Nuclear Reactors Lead Storage Batteries for Nuclear Power Plants RG Division Title Rev. Status Conformance Comments Section 1.160 of Effectiveness the Monitoring 1.152 for Use of Computers in Criteria 1.1531.155 for Safety Systems Criteria Station Blackout 11.156 11.157 Conforms Qualification of Connection Applicable to EDSS. Calculations of Best-Estimate Partially Conforms Thedesign 1.159 Assuring the Availability of Funds 3.11 1.158 Qualification of Safety-Related
Tier 2 1.9-25 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable frequenciesOption of A or aling becomes necessary, the tests that are theresponsibilityCOL applicant of the or licensee. quake planning and post-earthquakequake planning actions) that and are theresponsibilityof the COL applicant or licensee. performance of local leak rate tests (Type B and Type with the guidance provided in in accordance tests) C RegulatoryANSI/ANS 56.8, 94- NEI and Guide 1.163, 01. The NuScale system design, in conformance with 10 accommodates the 10 CFR 50.54(o), 50, CFR method Appendix J, test Option B. This RG is the responsibilityof a COL appli- cant or licensee that seeks to implement OptionB. tor vessel materials will exceed the 50 ft-lb energy signifi-withof the vessel life value (throughout the this guidancecant margin) below whichwould the reactor apply.in the unlikely event However, vessel material surveillance program implemented during reactor operations indicates that this is not the case, the requirements of 10 CFR 50, Appendix G would be the respon- 1.161 RG of the provisions and (see discus- sibility of the COL applicant or licensee 1.162). RG of sion of provisions and the10 CFR 50.66 requirements of RG 1.162 would be theresponsibilityof the COL applicant or licensee. - Applicable Not This RG governs post-earthquakeinspections and - Applicable Not This RG governs programmatic activities (earth- -- Applicable Not If thermal anne Applicable Not penetrations Thedesign of containment supports - Applicable Not TheCharpyreac-the NuScale upper-shelfenergy of Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table clear Power Plant Shut Down by a Seismic Event Shut Immediate Nuclear Power Plant Immediate Actions Operator Post Earthquake Thermal Annealing of Reactor Pressure Vessels Leak-Test Program Vessels with Charpy Upper-Shelf Upper-Shelf Charpy with Vessels Energy Less Than50 Ft-Lb RG Division Title Rev. Status Conformance Comments Section 1.167 Restart of a Nu 1.166and Planning Pre-Earthquake 1.162 Format and Content of Report for 1.163 Performance-Based Containment 1.161 Reactor Pressure of Evaluation
Tier 2 1.9-26 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.2 7.2 7.2 7.2 7.2 14.3 14.3 14.3 14.3 14.3 cycle, which is differ-cycle, n to some administrative n IEEE 829-2008 are tailored to the are 829-2008 IEEE exception to some administrative exception to some administrative e requirements of IEEE 828-2005 are 828-2005 requirements of IEEE e mandatory requirements in IEEE 830-1998 standard 830-1998 requirements in IEEE mandatory that conflict with established Engineering or quality documentation requirements. mandatory requirements in IEEE 1008-1987 that that 1008-1987 requirements in IEEE mandatory conflictEngineering or quality doc- with established umentation requirements. NuScale I&C development life in lifecycle from the conceptual waterfall ent RG 829-2008 IEEE tasks from The applicable 1.152. are mapped to the NuScale I&C development lifecy- cle. NuScale takes exceptio that con- mandatory requirements in the standard flict with established Engineering or quality docu- mentation requirements. IEEE Std. 7-4.3.2-2003 that it endorses. For RG 1.168, endorses. For RG 1.168, that it 7-4.3.2-2003 Std. IEEE to are tailored 1012-2004 requirements of IEEE the I&C developmentNuScale theis dif- lifecycle, which in lifecycle waterfall conceptual from the ferent IEEE applicable tasks from The 1012-2004. IEEE development are to the I&C 1012-2004 mapped. Some administrative mandatory require- ments in the standard conflict with established Engi- documentationrequirements.neering or QA The are tailored to the 1028-2008 IEEE requirements of NuScale I&C development lifecycle. tailored development to the NuScale I&C lifecycle, differentiswhich from the conceptual waterfall life- applicable tasks from The cycle in RG 1.152. IEEEthe NuScale I&C devel- are mapped to 828-2005 opment lifecycle. 1 Partially Conformstakes NuScale 1 Partially Conformstakes NuScale 1 Partially Conforms For this RG, th 1 Partially Conformsof Requirements 2 Partially Conforms The NuScale design applies Revision 3 and RG 1.152, Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table fety Systems of of Systems fety tionsDigital Computer Soft- for ware Used in Sa Nuclear Power Plants Computer Software Used in Software Computer Safety Systems of Nuclear Power Plants for Digital Computer Software Software Digital Computer for Used in Safety Systems of Nuclear Power Plants Used in Software Computer Safety Systems of Nuclear Power Plants and Audits for Digital Computer Digital Computer for Audits and SoftwareSystems Used in Safety Nuclear Power Plants of RG Division Title Rev. Status Conformance Comments Section 1.172 Software Requirements Specifica- 1.171 Unit Testing for Digital Software 1.169 Configuration Management Plans 1.170 Test Documentation for Digital 1.168 Verification, Validation, Reviews,
Tier 2 1.9-27 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.2 16.1 14.3 Not Applicable Not Applicable Not Applicable Not Applicable fecycle, which differs quality documentation quality specific operational pro- administrative mandatory sk-informed approach for le to licensees seeking change to le to licensees verns site-specificdecommission- IEEE 1074-2006 are tailored to the are tailored to 1074-2006 IEEE ing and license termination planning anding and license termination planning imple- mentation activities that are the responsibility of the COL applicant or licensee. approvalplant-specificto their of changes technical specifications.considered this However, NuScale guidance, as appropriate, in risk-informed technical specification development. informed inservice inspection program for piping. Such a program is a plant- responsibilitygram that is theof the COL applicant or licensee. licensing basis using a risk-informed approach and is theresponsibilityof the COL applicant or licensee. not using a ri is NuScale IST. licensingand basis, is the responsibility the of licensee. NuScale I&C development li 1.152. lifecycle in RG waterfall conceptual from the Applicable are tasks from IEEE 1074-2006 mapped to I&C developmentNuScale the lifecycle. NuScale takes exception to some that conflict with requirements in the standard established Engineering or requirements. - Applicable Not This RG is applicab 1 Applicable Not This guidance go 1 Applicable Not 1 This RG applies to existing licensees seeking NRC Not Applicable This RG addresses the use of PRA in support of a risk- 1 Partially Conforms of Requirements 2 Not Applicable This RG is applicable to licensees seeking changes in Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table ware Life Cycle Cycle ware Life License TerminationPlans for Nuclear Power Reactors Risk-Informed Decision making: making: Decision Risk-Informed Technical Specifications making Decision Risk-Informed Inservicefor Inspection of Piping Risk Informed Decision making: Decision Informed Risk Inservice Testing Processesfor Digital Computer SoftwareSystems Used in Safety Nuclear Power Plants of tic Risk Assessment in Risk-tic Risk Assessment in Informed Decisions on Plant-Spe- Changes to the Licensingcific Basis RG Division Title Rev. Status Conformance Comments Section 1.179Format and Content of Standard 1.177 Plant-Specific, Approach for An 1.178 Plant-Specific Approach for An 1.175 Plant-Specific, Approach for An 1.173 Soft Developing 1.174 An ApproachProbabilis-Using for
Tier 2 1.9-28 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.4 9.3 7.2 9.5 12.2 15.6 15.7 3.11 15.0.2 15.0.3 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable NEI 97-04 Appendix B is the 97-04 NEI site-specific decommissioning plan- site-specific decommissioning plan- seeking to renew their operating licenses. thatapplicant are the responsibilityor the COL of licensee. responsibilitylicensee.or of the COL applicant ning activities that are the responsibility of the COL responsibility of the ning activities that are the applicant or licensee. ning and implementationning and activitiesthat are the responsibilitylicensee.or of the COL applicant that coredamage does not occur duringdesign any basis event. Thus, the RG 1.183 guidance is partially applicable to the NuScale dose consequence analy- Thesis. basis and justification for departures from conse- the limiting dose for guidance RG 1.183 the quence analysisNuScale are provided in a Topi- for uses the alternative source term Report. NuScale cal non-LOCA or transient-specific guidance of RG 1.183 15 events. for Chapter tiesof the COL applicant thatresponsibility are the or licensee. to address effects of electromagneticaddressto effects and radio-fre- interference are applicable. (EMI/RFI) quency design of to the related guidance Aspects of this site-specificSSC and installation and testing prac- power tices and for the effects of EMI/RFI addressing surges on safety-related I&C systems are the respon- sibility of the COL applicant or licensee. - Applicable Not This RG implements change process requirements - Applicable Not This RG endorses - Applicable Not This guidance governs site-specificreporting activi- 1 Applicable Not This RG is applicable to operating reactor licensees 1 Applicable Not This RG governs 1 Applicable Not This RG governs 0 Partially ConformsNuScale For the design, the safety analysis shows 1 Partially Conforms Aspects of thisSSC guidance related to the design of Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Applications to Renew Nuclear Power PlantOperating Licenses 10 CFR Tests, and 50.59, Changes, Experiments tifying 10tifying CFR Design Bases 50.2 Post-Shutdown Decommission- ing Activities Report Power Reactors Terms for Evaluating DesignBasisTerms for Evaluating AccidentsNuclear Power Reac- at tors Safety Analysis Report in Accor- Report Analysis Safety dance with 10 CFR 50.71(e) magnetic and Radio-Frequency Interference in Safety-Related InstrumentationSys- and Control tems RG Division Title Rev. Status Conformance Comments Section 1.188Format and Content for Standard 1.187 Guidance Implementation of for 1.186 for Iden- Examples and Guidance 1.185Format and Content for Standard 1.184 of Nuclear Decommissioning 1.183Source Radiological Alternative 1.181 Content of the Updated Final 1.180 Electro- for Evaluating Guidelines
Tier 2 1.9-29 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.2 5.3 9.4 9.5 11.3 14.3 19.1 15.0.3 Appendix 9A Not Applicable Not Applicable Not Applicable idance of RG 1.190 with rsuant to 1050.55a(z). CFR r new reactors and existing r new reactors and le except for portions (1) directed portions (1) le except for and fluence calculation methods methods calculation fluence and licensees of existing reactors thatlicensees are not authorized of under use the alternative source term to 10 CFRto be is specified 1.183 Therefore, RG 50.67. fo RG 1.195 used in lieu of the alternativereactors authorized source to use under 10term CFR 50.67. authorized by the NRC pu bility of the COL applicantofbility the COL or licensee. activities that are applicablereac-holders only to of licensestor that have permanently ceased power operations. exceptions as described in NuScale Technical Report "Fluence Calculation Methodology TR-0116-20781, and Results." are consistent with the gu are consistent with the toward a specific reactor design (e.g., BWR or non- BWR (e.g., specific reactor design a toward the NuScale relevant to not or SSC conditions LWR) PWR design; andrelated (2) to site-specific fire pro- tection systems and equipment or programmatic and procedural activities that are the responsibility of the COL applicant or licensee. CnomNn.9.4 -ConformsNone. - Applicable Not by source terms used This RG pertains to TID14844 - Not Applicable This RG governs site-specific fire protection program - Conforms Partially neutron flux The 4 Conforms cases in RGcode ASME unless are not used 1.193 1 Applicable Not is the responsi- Implementation of in-service testing 2 Partially Conforms This RG is applicab Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table tions for Control Room Radiologi- Control Room tions for cal Habitability Assessmentsat Nuclear Power Plants Evaluating RadiologicalConse- quences of DesignBasisAcci- Light-Water Nuclear dents at Power Reactors for Use Code Case Acceptability, ASME Code OM Code Nuclear Power Plants During andDecommissioning Perma- nent Shutdown Plants Methods for Determining Pres- Methods for sureVessel Neutron Fluence RG Division Title Rev. Status Conformance Comments Section 1.194Concentra- Relative Atmospheric 1.195for Methods and Assumptions 1.193Cases Not Approved Code ASME 1.192 Operation and Maintenance 1.191 Fire Protection for Program 1.189Protection for Nuclear Power Fire 1.190 and Dosimetry Calculational
Tier 2 1.9-30 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.8 3.8 6.4 19.1 18.7 (via the 18.7 (via Not Applicable Not Applicable Human-System Human-System Summary Report) Interface Design Results e DCA. References to ESF ed using compressed air 0 with regard to PRA quality 0 with regard to quality PRA rned by this guidance is the guidance is applicable but the spe- and technical adequacy. cific language endorses Appendix B of ACI 349-2001 349-2001 ACI Appendix B of endorses cific language the area of loadin combi- with specified exceptions version of the ACI 2006 NuScale uses the nations. standard. 349 responsibilitylicenseeor of the COL applicantrefer- certifiedencing the design. the COLthe applicant or licensee referencingthe certi- design. fied ity design within the scope of the standard plant the scopewithinity design of to th applicable are design to the NuS- ventilation systems are not applicable habitability NuScale control room design. The cale filtra- neither relies on nor uses emergency system tion to protect operators during accident conditions. Rather, clean air is provid of this guidance specify opera-tanks. Other aspects tional, programmatic activities that are the responsi- applicantofbility the COL or licensee. - Partially Conforms intent of this The - Not Applicable The evaluation gove - Applicable Not arethe responsibilitytesting activities Inleakage of 2 Conforms As19. referenced in SRP 1 Partially Conforms Aspects of this habitabil- RG related to control room Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Soil Liquefac- Technical Adequacy of Probabilis- tic Risk Assessment Results for Risk-Informed Activities Structural Supports in Concrete Supports Structural Assessing Seismic Power Plant Sites at Nuclear tion Envelope Integrity at Nuclear Nuclear at Integrity Envelope Power Reactors Light-Water Nuclear Power Reac- tors RG Division Title Rev. Status Conformance Comments Section 1.200for Determining the Approach An 1.199 and Components Anchoring 1.198Criteria for Procedures and 1.197Room Control Demonstrating 1.196 at Room Habitability Control
Tier 2 1.9-31 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.0 7.2 8.1 8.3 3.2 Not Applicable to that prescribed in to ogy is consistent with of deterministic engineer- the traditional the traditional approach decommissioning cost estimates that are applicable only to licensees. framework for a licenseeframeworkuse a risk-informed to pro- cess for categorizing SSC by their safety significance, based on thisand process can remove SSC of low safety significanceof identified the scope from spe- treatment requirements. Thus, these require- cial ments are applicableto licenseeschoosethis that uses a risk-informed,alternative framework. NuScale performance-based approachsafety to classification the strengthsthat blends ing judgment and probabilistic methods. Specifi- safety to SSC approach the NuScale cally, classification combines using the definitions of 10 CFR 50.2 and of guidance with the alternative and SRP Section 3.2.2 RG 1.26 similar regulatory framework 10 CFR00-04 endorsed 1.201 (and NEI and RG 50.69 This methodol RG 1.201). by recom- which and SECY-10-0034, SECY-03-0047 of a probabilistic, risk-informed mend the use approach for SSC safety classification.NuScale NEI and RG 1.201 of applies the guidance to 00-04 extentthe appropriate the baseline risk given met- reactorNuScale advancedrics for the design. CnomNn.3.8 --ConformsNone. Conforms None. 15.0.2 - Applicable Not This RG implements regulatory requirements for 1 Partially Conforms 10 CFR provides an alternative regulatory 50.69 Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Methods tionof Nuclear Power Plants Decommissioning Cost Decommissioning Estimates Nuclear Power Reactors for tures, Systems, and Components Power Plants Accord- in Nuclear ing to Their Safety Significance RG Division Title Rev. Status Conformance Comments Section 1.2031.204and Transient Accident Analysis Protec- for Lightning Guidelines 1.202Format and Content of Standard 1.201Struc- Guidelines for Categorizing
Tier 2 1.9-32 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.3 7.2 3.9 14.3 3.12 Appendix 3C Appendix Not Applicable Not Applicable Not Applicable Ch. 1 through 19 gulatory positions in RG gulatory positions in d. 485-1997, with consider- 485-1997, d. DC power systems conform to the VRLA VRLA to the conform power systems DC sizing guidance in IEEE St guidance in IEEE sizing ation as appropriate for re battery sizing. relevant to VRLA 1.212 tery battery chargers are chargers or inverters; EDSS not located in a harsh environment. exceptions. ground motion response spectra and is therespon- sibility of the COL applicant or licensee. that are developing or revising a risk-informed, per- fire protection program pursuant formance-based to 10 CFR Development 50.48(c). and implementa- tion of a risk-informed, performance-based fire pro- tectionprogramwould be the responsibilityCOL of applicants or licenseesthe NuScale that reference provisions elect to implement theand that design, of 10 CFR 50.48(c). - Conforms None. 3.11 - Applicable Not bat- safety-related TheNuScale design does not use - Conforms None. 3.11 CnomNn.3.8 --ConformsNone. Partially Conforms This RG is thetemplate for theFSAR layout, with - Applicable Not Thisof guidance site-specific is for development 1 Applicable Not This RG applies to reactorlicensees or applicants November 2008 Partially ConformsNuScale The Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Cables and Field Splices for and Field Splices Cables Nuclear Power Plants age Batteries Battery Chargers and Inverters for Chargers and Inverters Battery Nuclear Power Plants Qualification of Safety-Related Computer-Based Instrumenta- tionandin Control Systems Nuclear Power Plants for Nuclear Power Plants (LWR for Edition) the Life Incorporating Analyses Components Reduction of Metal Light- of the the Effects to Due Environment for Water Reactor New Reactors Site-Specific Earth- Define the to quake Ground Motion Based Fire Protection for Existing Protection for Existing Based Fire Light-Water Nuclear Power Plants RG Division Title Rev. Status Conformance Comments Section 1.2111.212 Qualification of Safety-Related of Sizing Large Lead-Acid Stor- 1.210 Qualification of Safety-Related 1.209 Guidelines for Environmental 1.2061.207 License Combined Applications Fatigue for Evaluating Guidelines 1.208 Performance-Based Approach A 1.205 Performance- Risk-Informed,
Tier 2 1.9-33 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.3 3.5 3.8 6.2 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable al system design does notsystem design use al evaluation activities thata are the responsibility of license or constructionpermit applicant. environmental monitoring activities that are the responsibilitylicensee.or of the COL applicant Regulatory Position 1 (which occurs in Figure 2 of RG design basis wind speed for the 0) as the Rev. 1.221 hurricane. Confirmation of sitecharacteristics is the responsibility of the COL applicant. reactor licensees, including COL holders. reactor licensees, including ject to conditionject to monitoring during the develop- of the maintenance rule (10ment CFR 50.65) inclusion the criteria for program. Cables that meet in the maintenance rule program are subject to the guidance of RG 1.218. complying with the requirements of 10 with thecomplying CFR 52.99, applicablewhichis applicants to COL and licensees. safety-related motor control centers. - Conformsspeed postulatedthe highest wind in uses NuScale - Applicable Not Theseareapplicable requirements to operating -- Conforms Applicable Not None. TheCOL holder determines whether a cable is sub- 19.5 - Applicable Not TheNuScale electric 2 Applicable Not This guidance governs site-specificenvironmental 2 Applicable Not This guidance governs site-specific,programmatic CnomNn.3.8 0ConformsNone. 12 Conforms Applicable Not None. This guidance describes acceptable methodsof 19.5 Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Reports for Nuclear Power Sta- Reports tions itoring for Nuclear Power Plants Power itoring for Nuclear cane Missiles for Nuclear Power Plants Emergency Plans for Nuclear Power Reactors Evaluation for Internal Pressure Evaluation Loadings Above Design-Basis Pressure Beyond-Design-Basis Aircraft Impacts niques for Electric Cables Used in Nuclear Power Plants Aircraft Threats Under 10 CFR Part 52 Motor Control Centers for for Nuclear Motor Control Centers Power Plants RG Division Title Rev. Status Conformance Comments Section 4.2Environmental of Preparation 4.1Mon- Environmental Radiological 1.221 Hurricane and Hurri- Design-Basis 1.219Changes to on Making Guidance 1.216 Structural Containment Integrity 1.2171.218 of the Assessment for Guidance Condition-Monitoring Tech- 1.2141.215 for Response Strategies Potential Closure ITAAC for Guidance 1.213 Qualification of Safety-Related
Tier 2 1.9-34 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable systems. These activitiessystems. These able only to uranium mills.able only to uranium Not Applicable icable only to fuel cycle facili- cycle fuel icable only to ns site-specific evaluation activi- ies only to uranium enrichment ies only to applicable to licensees seeking applicable to licensees itories. ities that are the responsibilityities that are the of a COL applicant or holder. ties. evaluation activities thata are the responsibility of license or constructionpermit applicant. tiesof the COL applicant. thatresponsibility are the facilities. activities related to modeling temperature impact of modeling to activities related plant discharge on aquatic are theresponsibilityof the COL applicant or licensee. renewal of their operating license.renewal of - Applicable Not This guidance governs site-specificenvironmental 12 Applicable Not This guidance is applic Applicable Not 2 Applicableapplicants. to COL 1 Applicable Not This guidance is appl Applicable Not This guidance is applicable only to geological repos- Not Applicable 1 Applicable Not This guidance governs site-specificproceduralactiv- 12 Applicable Not This guidance appl Applicable Not This guidance governs site-specificenvironmental 2 Not Applicable This guidance gover 1 Applicable Not This guidance is Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table ability Criteria for mental Monitoring at Uranium Mills cal MonitoringPrograms (Incep- tion through Normal Operations License Termination) - to Effluent andStreams the Environment active Materials in Liquid and Gaseous Effluents fromNuclear Fuel Cycle Facilities Site Characterization Plans for High-Level-Waste Geologic Repositories dural Specifications for Thermolu- minescence Dosimetry: Environmental Applications Reports for Commercial Uranium for Commercial Uranium Reports Enrichment Facilities Nuclear Power Stationsfor Nuclear Power Stations matical ModelsSelected Pre- to Dispersion in Effluent Heated dict Bodies Water Natural tion of Supplemental Environ- for Applications mental Reports Plant Power Renew Nuclear to Operating Licenses RG Division Title Rev. Status Conformance Comments Section 4.144.15Environ- and Effluent Radiological Quality Assurance for Radiologi- 4.16 Monitoring and Reporting Radio- 4.17Format and Content of Standard 4.13Testing, and Proce- Performance, 4.94.11Environmental of Preparation Terrestrial Environmental Studies 4.7 Site Suit General 4.4 Reporting Procedure for Mathe- 4.2S1 Prepara- 1 to RG 4.2, Supplement
Tier 2 1.9-35 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.1 9.2 9.4 10.4 11.2 11.3 12.3 12.5 14.3 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable towards licensees oftowards fuel fabrica- licensees towards licensees oftowards fuel process- licensees pplicable only to near-surface dis- near-surface pplicable only to pplicable only to near-surface dis- near-surface pplicable only to ance is applicable only to non-reactor facili- tion facilities. facilities. ing and fuel fabrication facilities. that relate to site-specific, operational aspects that aspects that to site-specific, operational relate that are theresponsibilityof the COL applicant or design. licensee referencing the NuScale ties. posal of radioactive waste, which is not within the posal of radioactive the NuScalecertified design. scope of posal of radioactive waste, which is not within the posal of radioactive the NuScalecertified design. scope of - Applicable Not This RG is directed - Applicable Not Thisoftowards enrichment RG is directed licensees -- Applicable Not This RG is applicable to operating reactor licensees. Applicable Not This RG is directed Not Applicable - Partially Conforms This guidance is applicable,the portions except for - Applicable Not This guidance is a - Applicable Not This guidance is a 1 Applicable Not guid This Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Mass Spectrometric, and Spectro- chemical Analysis of Nuclear- Dioxide Powders Grade Uranium and Pellets the Measurement of Uranium Uranium and (UF4) Tetrafluoride (UF6) Hexafluoride During Operations Materialsfor Special Nuclear tion Accountability and Control and Radioactive Waste Genera- Life-Cycle Planning tion: borne Radioactive Materials to borne Radioactive the Environment for Licensees than Power Reactors other Near-Surface Disposal of Low- Level Radioactive Waste Environmental Reports for Near- Reports for Environmental Surface Disposal of Radioactive Waste RG Division Title Rev. Status Conformance Comments Section 5.5 for Methods Chemical, Standard 5.4 Analytical Standard Methods for 4.225.3 Planning Decommissioning Terminology and Nota- Statistical 4.21 Minimization of Contamination 4.20 on Constraint Releases of Air- 4.19 Sites for Selecting for Guidance 4.18Format and Content of Standard
Tier 2 1.9-36 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria nical Report) nical Report) Not Applicable Not Applicable Not Applicable Not Applicable 13.6 (via Security Tech- 13.6 (via Security Tech- this guidance this guidance le to a specialmaterial nuclear ogrammatic aspects of ogrammatic aspects of cale is not a special nuclear material licensee. licensee. are theresponsibilityof the COL applicant or design. licensee referencing the NuScale NuScale design does not process SNM. equipment, and methods that are not applicable to are not applicable methods that and equipment, the NuScale design. are theresponsibilityof the COL applicant or design. licensee referencing the NuScale - Applicable Not Applicablefacilities. to Part 70 Not Applicable - Applicable Not - Applicableapplicant to COL or licensee. Applicable Not ThisDCA because NuS- RG is not applicableto the Not Applicable -- Applicable Not Applicablefacilities. to Part 70 Applicable Not This RG is applicab Not Applicable - Partially Conformspr Site-specific, 1 Applicable Not 1 Applicablefacilities. to Part 70 Applicable Not Applicablefacilities. to Part 70 Not Applicable Not Applicable 1 Applicable Not Applicablefacilities. to Part 70 Not Applicable 1 Applicable Not This RG applies to facilities that process SNM. The 12 Applicable Not Applicablefacilities. to Part 70 Applicable Not This guidance governs processes,procedures, Not Applicable 1 Partially Conformspr Site-specific, Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table ual Holdup ual mizing Residual Holdup of Special for Material in Equipment Nuclear Wet Process Operations Areas ControlAreas and Item ing of Guards and Watchmen ing Ray Enrichment Assay by Gamma Spectrometry Normality (Employing Individual Observed Values) ical Inventories of Calculation in Nuclear ciples Materials Control tection and Controls of Facilities and SpecialMaterials Nuclear Nuclear Material Contained in Nuclear Material Contained ScrapWaste and mizing Residual Holdup of Special and Nuclear Material in Drying Fluidized Bed Operations troscopy Systems for Measure- Nuclear Material ment of Special Areas, Vital Areas, and Material Access Areas RG Division Title Rev. Status Conformance Comments Section 5.235.25Resid- Situ Assay of Plutonium In Considerations Design for Mini- 5.26 Selection of Material Balance 5.205.21Qualify- Training, Equipping, and 5.22 Nondestructive Uranium-235 Assessment of the Assumption of 5.135.18 Phys- Material Nuclear of Conduct and Prin- Error Concepts Limit of 5.12 Use of Locks in the Pro- General 5.11 Nondestructive Assay of Special 5.85.9Considerations Design for Mini- Guidelines for Germanium Spec- 5.7 Control Entry/Exit for Protected
Tier 2 1.9-37 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria nical Report) Not Applicable Not Applicable Not Applicable Not Applicable 13.6 (via Security Tech- this guidance le to a specialmaterial nuclear icable to but the NuScale design ogrammatic aspects of are theresponsibilityof the COL applicant referenc- ing the NuScale design. cale is not an applicant for a specialmaterial nuclear license. form in an unsealed licensee. tion licensees. the mate- to meet a COL applicant used by be may rial control and accounting requirements in Subpart B of 10 CFR Part 74. -- Applicable Not Applicableapplicant to COL or licensee. Applicable Not Issued for comment. Not Applicable Not Applicable - Applicable Not Applicablefacilities. to Part 70 Not Applicable - Applicable Not ThisDCA because NuS- RG is not applicableto the - Applicable Not Applicablefacilities. to Part 70 Not Applicable -- Applicable Not Applicableapplicant. to COL Applicable Not This RG appliesfuel fabrica- to fuel processing and Not Applicable 3 Partially Conformspr Site-specific, 21 Applicable Not This RG is not appl Applicable Not 1 Applicableapplicant. to COL 1 Applicable Not Applicableprocessing. to Part 70 Applicable Not This RG is applicab Not Applicable Not Applicable Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table for Measuring the Mass of Liquids Mass the Measuring for Nuclear Material (for Comment) mizing Residual Holdup of Special for Material in Equipment Nuclear Dry Process Operations tems of Uranyl Nitrate Solutions for of IsotopicAssay, Distribution, and Impurity Determinations for Nuclear Power Plants for Armed Guards for Road Shipment Special Nuclear Material of For Unaccounted nium in Scrap Material by Sponta- Detection neous Fission with Outlyinging Observations Differences in the Transfer of Spe- cial Nuclear Materials way Monitors RG Division Title Rev. Status Conformance Comments Section 5.485.49 Considerations-Systems Design Special of Transfers Internal 5.42Considerations Design for Mini- 5.435.44Security Force Duties Plant Sys- Alarm Intrusion Perimeter - Applicable Not applicant Applicable to COL or licensee. Not Applicable 5.39 Methods for the Analysis General 5.29 Nuclear Material Control systems 5.315.33Vehicle with Specially Designed 5.34 of Material Evaluation Statistical 5.36 Nondestructive Assay for Pluto- Practice for Deal- Recommended 5.28 Evaluation of Shipper-Receiver 5.27Door- Nuclear Material Special
Tier 2 1.9-38 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable rial. tion features and security program activities that are theresponsibilityof the COL applicant or licensee. - Applicable Not Applicableapplicant to COL or licensee. Not Applicable 11 Applicable Not Applicableapplicant to COL or licensee.1 Applicable Not Applicableapplicant to COL or licensee. Applicable Not Not Applicable Applicableapplicant to COL or licensee. Not Applicable Not Applicable 00 Applicable Not Applicable to fuel cycle facilities. Not Applicable Applicable to transportation of special nuclear mate- Not Applicable 03 Applicable Not This RG applies to fuel cycle facilities. Applicable Not 1 Not applicablenuclear to power plants.1 Applicable Not Not Applicable This RG applies to fuel processing licensees. Not Applicable Applicable Not This guidance governs site-specificphysical protec- Not Applicable Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table of Strategic Special Nuclear Mate- Special Nuclear Strategic of rial Traceability of Special Nuclear Measure- Accounting Material ments a Licensee Physical SecurityPlan for the Protection of Special Nuclear Material of Moderateor Significance Low Strategic a Licensee Physical Protection Plan for Strategic Special Nuclear Material in Transit Safeguards Contingency Plans for Contingency Safeguards Fuel Cycle Facilities Plans for Contingency Safeguards Transportation Material Control and Accounting Material Control Systems a Licensee Physical Protection Plan for Strategic Special Nuclear Material at Fixed Sites (Other than Nuclear Power Plants) Estimation Methods for Error Nondestructive Assay Plans for Contingency Safeguards Nuclear Power Plants RG Division Title Rev. Status Conformance Comments Section 5.575.58Receiving Control and Shipping Considerations for Establishing 5.59Format and Content for Standard 5.60Format and Content of Standard 5.555.56Format and Content of Standard Format and Content of Standard 5.51Nuclear Review Management of 5.52Format and Content of Standard 5.535.54 Qualification, Calibration, and Format and Content of Standard
Tier 2 1.9-39 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria nical Report) nical Report) nical Report) nical Report) Not Applicable Not Applicable 13.6 (via Security Tech- 13.6 (via Security Tech- 13.6 (via Security Tech- 13.6 (via Security Tech- this guidance fuel cycle facilities. Applicable Not RG 5.71 that govern site-specific govern site-specific that 5.71 RG ogrammatic aspects of guidance appliessite-specific to security issues operationaldevel- activities (e.g., and programmatic opment and implementation of operational cyber applicantapplysecurity plans)the COL to or licensee. cant or licensee. licensee, thedesign must allow compliance(e.g., Layout Plan, which references Site parts of F090 73.55). program activities thatof are the responsibility the COL applicant or licensee. are theresponsibilityof the COL applicant or design. licensee referencing the NuScale concerning SNM and is the responsibilityconcerningSNM and of the COL applicant or licensee. - Not Applicable- COL applicant or licensee responsibility. Partially Conforms Theportions of Not Applicable - Applicable Not This guidanceresponsibility is theCOL appli- of the - Partially Conforms Although applicableapplicantorthe COL to -- Applicable Not Applicableapplicant to COL or licensee. Partially Conformspr Site-specific, Not Applicable - Applicable Not This guidance applies to 2 Applicable Not This guidance governs site-specificphysical security Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table the Theft and Diversion Design- the Theft and Basis Treat in the Design Develop- ament, and Implementation of Program that Security Physical (SGI) and 73.46 Meets CFR 73.45 Nuclear Facilities Radiological Sabotage Design- Threat in the Design, Devel- Basis opment and Implementation of a Program that Security Physical Meets 10 CFR 73.55 Require- ments (SGI) Use of Vehicles at Nuclear Power Vehicles at Use of Plants Nuclear Power Plants Shipments tectionPhysical of SecurityEquip- ment, and Key Lock Controls Protection Upgrade Rule Require- Protection Upgrade for Fixedments Sites RG Division Title Rev. Status Conformance Comments Section 5.70 Guidancethe Application for of 5.71for Security Programs Cyber 5.69 Guidancethe Application for of 5.68Malevolent Against Protection 5.66 Access Authorization Program for 5.635.65 for Transient Protection Physical Vital Area Access Controls, Pro- 5.61 the Physical of Intent and Scope 5.62 Reporting of Safeguards Events 1 Applicable Not This
Tier 2 1.9-40 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria nical Report) Not Applicable Not Applicable Not Applicable 13.6 (via Security Tech- fuels. Applicable Not ant or licensee to meet not use mixed oxide not use ation surveys and monitor- ation icable to but the NuScale design L applicant or licensee responsibility. Not Applicable activities related to radi activities related ing that are theresponsibilityof the COL applicant or licensee. Information against unauthorized disclosure in Information against unauthorized accordance with RG 5.79. tion program activitiesresponsibility that are the of COLthe applicant or licensee. may be used by a COL applicbe may requirementsfatigue management the of 10 CFR 26 Subpart I. -- Applicable Not ThisNuScale RG is not applicableto the design. - Not Applicable Not Applicable COL applicant or licensee responsibility. Not Applicable COL applicant or licensee responsibility. Not Applicable Not Applicable - Conforms Safeguards protects NuScale -- Not Applicable COL applicant or licensee responsibility. Applicable Not TheNuScale design does Not Applicable -- Applicable Not - This RG is not appl Not Applicable- COL applicant or licensee responsibility. Not Applicable COL applicant or licensee responsibility. Applicable Not This guidance governs site-specificphysical protec- Not Applicable Not Applicable 1 Applicable Not This guidance governs site-specific,programmatic Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Indicating DeviceSeals for Mate- rial Control and Accounting of Special Nuclear Material Development for Nuclear Power Reactors (OUO-SRI) Sites Construction Plant Power tion Surveys and Monitoring tion SRI) Oxide inFuels Nuclear Power Plants (SGI) Interface SecurityNuclear Personnel at Power Reactor Facilities Nuclear Power Reactors Power PlantPersonnel RG Division Title Rev. Status Conformance Comments Section 5.80 Pressure-Sensitive and Tamper- 5.815.83Set Identification and Target 5.848.2 Event Notifications Cyber Security for New Nuclear Fitness-For-Duty - Administrative Practices in Radia- Not Applicable CO 5.79Informa- of Safeguard Protection 5.775.78 Insider Mitigation Program (OUO- of Mixed Protection Physical 5.745.75 the Safety/Security Managing 5.76 Training and Qualification of Programs at Protection Physical 5.73 Fatigue Management for Nuclear
Tier 2 1.9-41 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.3 10.4 11.2 11.4 11.5 12.1 12.3 12.5 14.3 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable lection, maintenance,cali- this guidance is largely site-spe- activities, procedures, equipment, and methods that are theresponsibilityof the COL applicant. activities, procedures, equipment, and methods that are theresponsibilityof the COL applicant. apply to licensees for which uranium bioassay is apply to licensees for which uranium required. activities, procedures, equipment, and methods that are theresponsibilityof the COL applicant. COL applicant referencing the certified design.the cific and is the responsibilityis cific and of the COL applicant. NuScale However, the application for standard certificationdesign considered this guidance to be applicablenecessaryprovide to the extent to rea- sonable assurance that the COL applicant referenc- can meet these ing the certified design this guidancerequirements. The aspectsthat are of design-specific (i.e., pertain to design features, facili- ties,and equipment that are technically functions, relevant to the NuScale standard plant designe.g., - Position C.2) are applicablethe DCA. to activities related to recording and reporting dose reporting dose and to recording activities related COL applicant of the the responsibility are that data or licensee. activities related to the se to the activities related pocket dosimeters reading of bration, training, and thatapplicant are the responsibilityor the COL of licensee. - Applicable Not This guidance governs programmatic activities that 1 Applicable Not This guidance governs site-specific,programmatic 3 Applicable Not This guidance governs site-specific,programmatic 3 Partially Conforms Implementation of 1 Applicable Not This guidance governs site-specific,programmatic 2 Applicable Not This guidance governs site-specificprogrammatic 1 Applicable Not This guidance governs site-specificprogrammatic 1-R Applicable Not These site-specific aspects are the responsibility of Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table tory Protection Radiation Exposure taining Occupational Radiation Exposures as Low as Is Reason- ably Achievable nium that Occupational Radiation Exposures at Nuclear Power Sta- tions WillReason-Is Be as Low ably Achievable for a Assumptions and Equations, Bioassay Program Reporting Occupational Radia- tion Dose Data Direct-Reading Pocket Dosime- ters RG Division Title Rev. Status Conformance Comments Section 8.15 Programs for Respira- Acceptable 8.13 Instruction Concerning Prenatal 8.10 Operating Philosophyfor Main- 8.11 ApplicationsBioassay of for Ura- 8.8 Ensuring to Relevant Information 8.9 AcceptableModels, Concepts, 8.7 and for Recording Instructions 8.4Monitoring Device Personnel -
Tier 2 1.9-42 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 12.4 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable fabricate fuel with uranium fuel with uranium fabricate is guidance governs site-specific,programmatic pplicable only to uranium mills. Not Applicable training programs, plans,that are and procedures theresponsibilityof the COL applicant or licensee. activities, procedures, equipment, and methods that are theresponsibilityof the COL applicant or licensee. facilities that process or isotope. the U-235 enriched with the workplace in to air sampling activities related thatapplicant are the responsibilityor the COL of licensee. medical institutions. plants. that relate to site-specific, operational aspects that aspects that to site-specific, operational relate that are theresponsibilityof the COL applicant referenc- ing the NuScale design. Construction activitiesdose appli-assessmentsthe COL are the responsibility of cant or licensee. activities, procedures, equipment, and methods that are theresponsibilityof the COL applicant or licensee. medical institutions. - Applicable Not This guidance governs site-specificoperational - Applicable Not This guidance governs site-specific,programmatic 2 Applicable Not This guidance governs activities applicable only to 1 Applicable Not This guidance governs activities applicable only to 1 Applicable Not Applicable onlymanufacturing to processing and 1 Partially Conforms This guidance is applicable,the portions except for 2 Applicable Not This guidance governs site-specific,programmatic 2 Applicable Not This guidance governs activities applicable only to Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Personnel at Light-Water-Cooled Personnel at Light-Water-Cooled Nuclear Power Plants sion and Activation Products Enriched Uranium-235 Processing Uranium-235 Processing Enriched and Fuel Fabrication cal Institutions Byproduct Material at NRC Material at NRC Byproduct LicensedProcessing and Manu- facturing Plants AssessmentReac- in Light Water tor Power Plants - Design Stage Man-Rem Estimates Radioiodine that Radiation Exposures at Medi- Be cal Institutions Will as Low is Reasonably Achievable RG Division Title Rev. Status Conformance Comments Section 8.27 Radiation ProtectionTraining for 8.25the Workplacein Sampling Air 8.26 ApplicationsBioassay of for Fis- 1 Not Applicable Th 8.24 Physics Surveys During Health 8.23Medi- at Radiation Safety Surveys 8.22at Uranium Mills Bioassay 1 Not Applicable A 8.21 Physics Surveys for Health 8.19 Occupational Radiation Dose 8.20 ApplicationsBioassay of for 8.18 Relevant to Ensuring Information
Tier 2 1.9-43 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 12.1 12.3 12.5 14.2 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable idance to beidance to applicable guidance governs site-specific,programmatic is the responsibility of the COL applicant.is the responsibilityof the COL However, considered this gu NuScale reasonable assur- to provide necessary the extent to ance that the COL applicant or licensee referencing the certified design can meet these requirements. materials facilities. materials activities, procedures, equipment, and methods that are theresponsibilityof the COL applicant or licensee. activities, procedures, equipment, and methods that are theresponsibilityof the COL applicant or licensee. activities, procedures, equipment, and methods that are theresponsibilityof the COL applicant or licensee. activities, procedures, equipment, and methods that are the responsibility of a licensee authorized to pos- sess nuclear material. uranium recovery facilities. uranium recovery facilities. training and instructional activities that are the responsibilitylicensee.or of the COL applicant activities, procedures, equipment, and methods that are theresponsibilityof the COL applicant or licensee. - Applicable Not This guidance governs activities applicable only to - Applicable Not This guidance governs site-specific,programmatic - Applicable Not This guidance governs site-specific,programmatic - Applicable Not This guidance governs site-specific,programmatic 1 Partially Conforms Implementation of this guidance is site-specific and 1 Applicable Not This guidance governs activities applicable only to 1 Applicable Not This guidance governs activities applicable only to 1 Applicable Not This guidance governs site-specific,programmatic Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Very High Radiation Areas in Nuclear Power Plants Materials Facilities Fetus to Calculate Occupational Radia- tion Doses Bioassay Program that Occupational Radiation Exposures at Uranium Recovery BeasFacilities Will Low as Is Rea- sonably Achievable nium Recovery Facilities nium Recovery Occupational Radiation Exposure RG Division Title Rev. Status Conformance Comments Section 8.38High and to of Access Control 8.37 from for Effluents Levels ALARA 8.36 Radiation Dose to the Embryo/ 8.35Special Planned Exposure 1 Applicable Not This guidance governs site-specific,programmatic 8.34 Monitoring Criteria and Methods 8.32Tritiuma for Establishing Criteria 8.31 Relevant to Ensuring Information 8.30 Physics Surveys in Ura- Health 8.29 Instruction Concerning Risks from 8.28 Audible-Alarm Dosimeters - Applicable Not This
Tier 2 1.9-44 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable verns dosimetry methods for deter- mining effective dose equivalent for external radia- tion exposures. These methods are the responsibility of the COL applicant or licensee. facilities that administer radio-pharmaceuticals.facilities that - Not Applicable This guidance go - Applicable Not This guidance governs activities applicable only to Table 1.9-2: Conformance with Regulatory Guides (Continued) Regulatory with Conformance 1.9-2: Table Dose Equivalent from External Equivalent Dose Exposure Radioactive Materials RG Division Title Rev. Status Conformance Comments Section 8.40Measuring Effective Methods for 8.39 Release of Patients Administered
Tier 2 1.9-45 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 2.0 Table 2.0-1 Table 2.0-1 Not Applicable Not Applicable Not Applicable Not Applicable do not reference the do not reference the Comments Section ific Acceptance Criteria. Not Applicable This acceptance criterion is a pointerThis to other SRP sections. acceptance criterion is applicable only This COL applicantsto that DCA. acceptance criterion is for COL This applicants tomeet the design parameters in the DCA. established acceptance criterion is for COL This applicants tomeet the design parameters in the DCA. established acceptance criterion is applicable only This COL applicantsto that DCA. design parameters whereNuScale provides applicable. design parameters whereNuScale provides applicable. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms Ch 1 Conforms None. Ch 1 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Site-specific.Not Applicable Site-specific. Not Applicable Not Applicable Partially Conforms Partially Conforms Standard (DSRS) AC Title/Description Conformance SRP Acceptance Criteria SRP Acceptance Associated section SRP Each Referenced with Specific SRP Acceptance Criteria Contained in Related SRP Chapter 2 sections SRP Referenced Other or an Early Referencing Application COL Site Permit COL Application Referencing a Certified Design an Early Referencing Application COL Certified Design a Site Permit and COL Application Referencing Neither Nor a Certified Permit an Early Site Design and Design Qualification Testing Requirements Site Parameters Characteristics and Parameters and Design Design Characteristics Area Map and Personnel of or Removal Property, and Proposed and Activities Permitted AC II.1II.2 II.3 No Specific Acceptance CriteriaII.1 Performanceof New SafetyFeatures -II.2 II.3 Spec No II.4 II.5 App AApp A 1: Examples of Site Table All 2: Examples of Site-Related Table All Specification of Location and Site Exclusion Establishment of Authority, Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 1.0, Rev 2: Introduction Interfaces and SRP 1.0, Rev 2: Introduction Interfaces and SRP 1.0, Rev 2: Introduction Interfaces and Site 2.0, (March 2007): SRP CharacteristicsSite and Parameters Site 2.0, (March 2007): SRP CharacteristicsSite and Parameters Site 2.0, (March 2007): SRP CharacteristicsSite and Parameters Site 2.0, (March 2007): SRP CharacteristicsSite and Parameters Site 2.0, (March 2007): SRP CharacteristicsSite and Parameters Site 2.0, (March 2007): SRP CharacteristicsSite and Parameters Site 2.0, (March 2007): SRP CharacteristicsSite and Parameters Site Location Rev 3: 2.1.1, SRP Description and 2.1.2, Rev 3: Exclusion SRP Control and Area Authority
Tier 2 1.9-46 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 2.3.1 2.3.1 2.3.1 2.3.1 2.3.2 2.3.2 2.3.2 2.3.2 Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. Table 2.0-1 Table 2.0-1 Not Applicable Site-specific.Not Applicable Site-specific. Not Applicable Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Population Zone, Nearest Population Center Boundary, and Population Density Event Analysis Information Information AC AllAll Population Data, Exclusion Area, Low- All VariousAllIII.4.b.1 Event Probability and Design-Basis III.4.b.2 Various Postulated Site ParametersIII.4.b.3Tier 1 as Included Site Parameters III.4.b.4 TableSummary Site Parameters II.1 thru II.4 Not Applicable Conforms Basis for Site Parameters Site-specific. VariousIII.4.b.i None. ConformsIII.4.b.ii None. Postulated Site Parameters Not ApplicableIII.4.b.iii Site-specificTier 1 as Included Site Parameters ConformsIII.4.b.iv TableSummary Site Parameters None.All Conforms Basis for Site Parameters Not Applicable None. ConformsAll Not Applicable Site-specific. Various None.2.0-1 Table Conforms Various Not Applicable None.2.0-1 Table Table 2.0-1 Not Applicable Not Applicable Site-specific.2.0-1 Table Not Applicable2.0-1 Table Site-specific. Table 2.0-1 Not Applicable Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 2.1.3, Rev 3: Population 2.1.3, Rev 3: Population SRP Distribution Rev 3: 2.2.1-2.2.2, SRP Potential of Identification Vicinity Hazards in Site Evaluation of Rev 3: SRP 2.2.3, Potential Accidents 2.3.1, Rev 3: Regional SRP Climatology 2.3.1, Rev 3: Regional SRP Climatology 2.3.1, Rev 3: Regional SRP Climatology 2.3.1, Rev 3: Regional SRP Climatology 2.3.1, Rev 3: Regional SRP Climatology 2.3.2, Rev 3: Local SRP Meteorology 2.3.2, Rev 3: Local SRP Meteorology 2.3.2, Rev 3: Local SRP Meteorology 2.3.2, Rev 3: Local SRP Meteorology 2.3.2, Rev 3: Local SRP Meteorology 2.3.3, Rev 3: Onsite SRP Meteorological Measurements Program 2.3.4, Rev 3: Short-Term SRP Atmospheric Dispersion Estimates for Accident Releases
Tier 2 1.9-47 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None. 2.3.4 Conforms None. 2.3.4 Conforms None. 2.3.5 Standard (DSRS) (Continued) AC Title/Description Conformance Applicable Short-Term (Post- Applicable Short-Term LPZ, Parameters - EAB, Site Accident) and Control Room Atmospheric Dispersion Factors Information Information AC III (no number) III.6.b.1 Postulated Site ParametersIII.6.b.2Tier 1 as Included Site Parameters III.6.b.3 Conforms None. TableSummary Site Parameters III.6.b.4 Basis for Site Parameters Conforms None.All Conforms None.III.5.b.1 Various 2.3.4 Postulated Site ParametersIII.5.b.2Tier 1 as Included Site Parameters III.5.b.3 Conforms 2.3.4 None. TableSummary Site Parameters Not Applicable Site-specific. Conforms 2.3.4 None. Not Applicable 2.3.5 2.3.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 2.3.4, Rev 3: Short-Term SRP Atmospheric Dispersion Estimates for Accident Releases 2.3.4, Rev 3: Short-Term SRP Atmospheric Dispersion Estimates for Accident Releases 2.3.4, Rev 3: Short-Term SRP Atmospheric Dispersion Estimates for Accident Releases 2.3.4, Rev 3: Short-Term SRP Atmospheric Dispersion Estimates for Accident Releases 2.3.4, Rev 3: Short-Term SRP Atmospheric Dispersion Estimates for Accident Releases 2.3.5, Rev 3: Long-Term SRP Atmospheric Dispersion Estimates for Routine Releases 2.3.5, Rev 3: Long-Term SRP Atmospheric Dispersion Estimates for Routine Releases 2.3.5, Rev 3: Long-Term SRP Atmospheric Dispersion Estimates for Routine Releases 2.3.5, Rev 3: Long-Term SRP Atmospheric Dispersion Estimates for Routine Releases
Tier 2 1.9-48 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Applicable Site-specific. Not Applicable Status Review Plan (SRP) and Design Specific Review Conforms None. 2.4.2 Conforms None. 2.3.5 Conforms None. 2.4.1 Conforms None. 2.4.3 Standard (DSRS) (Continued) AC Title/Description Conformance rameters Summary Summary Tablerameters Conforms None. 2.4.2 e Parameters Included as Tier 1 as Included e Parameters Information Release) Site Parameters - Maximum Annual Average Site Boundary Atmospheric DispersionFactor Information Information AC III (no number)III (Routine Applicable Long-Term III.5.b.4 Basis for Site ParametersAll ConformsIII.7.B.i None.III.7.B.ii Various Postulated Site ParametersIII.7.B.iiiTier 1 as Included Site Parameters III.7.B.iv TableSummary Site Parameters Conforms Basis for Site Parameters None. Conforms None. Not ApplicableAll Site-specific. Conforms 2.3.5 None.III.4.B.i VariousIII.4.B.ii Postulated Site ParametersTier 1 as Included III.4.B.iii Site Parameters Conforms TableSummary Site Parameters 2.4.1 Not Applicable None. 2.4.1 Not Applicable Conforms Site-specific. None. 2.4.1 Not Applicable 2.4.3 2.4.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 2.3.5, Rev 3: Long-Term 2.3.5, Rev 3: Long-Term SRP Atmospheric Dispersion Estimates for Routine Releases 2.3.5, Rev 3: Long-Term SRP Atmospheric Dispersion Estimates for Routine Releases 2.4.1, Rev 3: Hydrologic SRP Description 2.4.1, Rev 3: Hydrologic SRP Description 2.4.1, Rev 3: Hydrologic SRP Description 2.4.1, Rev 3: Hydrologic SRP Description 2.4.1, Rev 3: Hydrologic SRP Description 2.4.2, Rev 4: FloodsSRP 4: Floods2.4.2, Rev SRP 4: Floods2.4.2, Rev SRP All III.11.B.i 4: Floods2.4.2, Rev SRP III.11.B.ii 4: Floods2.4.2, Rev SRP 2.4.3, Rev 4: Probable SRP Site Parameters Postulated III.11.B.iii Various Sit Maximum Flood (PMF) on III.11.B.ivStreams and Rivers Site Pa 2.4.3, Rev 4: Probable SRP Parametersfor Site Basis Maximum Flood (PMF) on Streams and Rivers Conforms 2.4.3, Rev 4: Probable SRP Maximum Flood (PMF) on None.Streams and Rivers 2.4.3, Rev 4: Probable SRP Maximum Flood (PMF) on ConformsStreams and Rivers Not None. 2.4.2 2.4.2
Tier 2 1.9-49 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 2.4.4 Conforms None. 2.4.5 2.4.6 Standard (DSRS) (Continued) AC Title/Description Conformance Information Information Information AC III.4.B.iv Basis for Site ParametersAllIII.8.B.iIII.8.B.ii Various Postulated Site ParametersIII.8.B.iii ConformsTier 1 as Included Site Parameters III.8.B.iv None. TableSummary Site Parameters All Conforms Basis for Site Parameters None. ConformsIII.7.B.i Various None. Not ApplicableIII.7.B.ii Postulated Site Parameters Site-specific. ConformsTier 1 as Included III.7.B.iii Site Parameters None. Conforms TableSummary Site Parameters III.7.B.iv 2.4.3 None. Basis for Site ParametersAll Not Applicable Conforms Site-specific.III.9.B.i None. 2.4.4 Not Applicable III.9.B.ii Various Postulated Site Parameters 2.4.4 III.9.B.iii ConformsTier 1 as Included Site Parameters None. TableSummary Site Parameters 2.4.4 Conforms None. Conforms Not Applicable 2.4.5 None. Not Applicable Site-specific. 2.4.5 2.4.5 2.4.6 Not Applicable 2.4.6 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 2.4.3, Rev 4: Probable 2.4.3, Rev 4: Probable SRP Maximum Flood (PMF) on Streams and Rivers 2.4.4, Rev 3: Potential SRP Dam Failures 2.4.4, Rev 3: Potential SRP Dam Failures 2.4.4, Rev 3: Potential SRP Dam Failures 2.4.4, Rev 3: Potential SRP Dam Failures 2.4.4, Rev 3: Potential SRP Dam Failures 2.4.5, Rev 3: Probable SRP Maximum Surge and Seiche Flooding 2.4.5, Rev 3: Probable SRP Maximum Surge and Seiche Flooding 2.4.5, Rev 3: Probable SRP Maximum Surge and Seiche Flooding 2.4.5, Rev 3: Probable SRP Maximum Surge and Seiche Flooding 2.4.5, Rev 3: Probable SRP Maximum Surge and Seiche Flooding 2.4.6, Rev 3: Probable SRP Maximum Tsunami Hazards 2.4.6, Rev 3: Probable SRP Maximum Tsunami Hazards 2.4.6, Rev 3: Probable SRP Maximum Tsunami Hazards 2.4.6, Rev 3: Probable SRP Maximum Tsunami Hazards
Tier 2 1.9-50 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None. 2.4.7 Conforms None.Conforms None. 2.4.8 Conforms None. 2.4.9 2.4.10 Standard (DSRS) (Continued) AC Title/Description Conformance Parameters Summary Summary TableParameters Conforms None. 2.4.7 Information Information Information Information AC III.9.B.iv Basis for Site ParametersAll ConformsIII.5.B.1 None.III.5.B.2 Various Postulated Site ParametersIII.5.B.3Tier 1 as Included Site Parameters III.5.B.4 TableSummary Site Parameters All Conforms Basis for Site ParametersIII.8.B.i None. ConformsIII.8.B.ii Various None. Postulated Site Parameters Not ApplicableIII.8.B.iii Site-specific.Tier 1 as Included Site Parameters ConformsIII.8.B.iv 2.4.6 TableSummary Site Parameters None.All Conforms Basis for Site ParametersIII.5.B.i None. ConformsIII.5.B.ii Various None. Postulated Site Parameters Not Applicable Site-specific.Tier 1 as Included Site Parameters Conforms 2.4.8 Not Applicable None. 2.4.8 Conforms None. 2.4.8 Not Applicable Site-specific. 2.4.9 Not Applicable 2.4.9 2.4.9 2.4.10 Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 2.4.6, Rev 3: Probable 2.4.6, Rev 3: Probable SRP Maximum Tsunami Hazards Effects 3: Ice 2.4.7, Rev SRP Effects 3: Ice 2.4.7, Rev SRP AllEffects 3: Ice 2.4.7, Rev SRP III.6.B.i III.6.B.iiEffects 3: Ice 2.4.7, Rev SRP Effects 3: Ice 2.4.7, Rev SRP III.6.B.iii Site Parameters Postulated Various 2.4.8, Rev 3: Cooling SRP Tier 1 as Included Site Parameters III.6.B.ivWater Canals and Reservoirs 2.4.8, Rev 3: Cooling SRP Site Water Canals and Reservoirs Parametersfor Site Basis 2.4.8, Rev 3: Cooling SRP ConformsWater Canals and Reservoirs 2.4.8, Rev 3: Cooling SRP None.Water Canals and Reservoirs 2.4.8, Rev 3: Cooling SRP Water Canals and Reservoirs Conforms2.4.9, Rev 3: Channel SRP Applicable Not Diversions None. Site-specific.2.4.9, Rev 3: Channel SRP Diversions 2.4.9, Rev 3: Channel SRP Diversions 2.4.9, Rev 3: Channel SRP Diversions 2.4.9, Rev 3: Channel SRP Diversions 2.4.10, Rev 3: Flooding SRP Protection Requirements 2.4.10, Rev 3: Flooding SRP 2.4.7 Protection Requirements Not Applicable 2.4.10, Rev 3: Flooding SRP Protection Requirements 2.4.7
Tier 2 1.9-51 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 2.4.11 Conforms None. 2.4.12 2.4.13 Standard (DSRS) (Continued) AC Title/Description Conformance Information Information Information AC III.5.B.iiiIII.5.B.iv TableSummary Site Parameters All Basis for Site ParametersIII.6.B.i ConformsIII.6.B.ii Various None. Postulated Site ParametersIII.6.B.iiiTier 1 as Included Site Parameters ConformsIII.6.B.iv TableSummary Site Parameters None.II.1 thru II.5 Conforms Basis for Site Parameters VariousIII.6.B.i None. ConformsIII.6.B.ii None. Postulated Site Parameters Not ApplicableIII.6.B.iii Site-specific.Tier 1 as Included Site Parameters ConformsIII.6.B.iv TableSummary Site Parameters None. 2.4.10 All Conforms Basis for Site Parameters Not Applicable None. Conforms Site-specific.III.5.B.i 2.4.10 Various None. Postulated Site Parameters Conforms 2.4.11 III.5.B.ii Not Applicable None. 2.4.11 Tier 1 as Included Site Parameters Conforms 2.4.11 None. Not Applicable Not Applicable Site-specific. 2.4.12 2.4.12 2.4.12 Not Applicable 2.4.13 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 2.4.10, Rev 3: Flooding SRP Protection Requirements 2.4.10, Rev 3: Flooding SRP Protection Requirements Water 2.4.11, Rev 3: Low SRP Considerations Water 2.4.11, Rev 3: Low SRP Considerations Water 2.4.11, Rev 3: Low SRP Considerations Water 2.4.11, Rev 3: Low SRP Considerations Water 2.4.11, Rev 3: Low SRP Considerations 2.4.12, Rev 3: SRP Groundwater 2.4.12, Rev 3: SRP Groundwater 2.4.12, Rev 3: SRP Groundwater 2.4.12, Rev 3: SRP Groundwater 2.4.12, Rev 3: SRP Groundwater 2.4.13, Rev 3: Accidental SRP Releases of Radioactive in Effluents Ground Liquid Waters and Surface 2.4.13, Rev 3: Accidental SRP Releases of Radioactive in Effluents Ground Liquid Waters and Surface 2.4.13, Rev 3: Accidental SRP Releases of Radioactive in Effluents Ground Liquid Waters and Surface
Tier 2 1.9-52 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None. 2.4.14 Standard (DSRS) (Continued) AC Title/Description Conformance Information AC III.5.B.iii TableSummary Site Parameters III.5.B.iv Basis for Site Parameters ConformsAll None.III.5.B.i Various Conforms Postulated Site Parameters None.III.5.B.iiTier 1 as Included Site Parameters III.5.B.iii Conforms TableSummary Site Parameters None. Not ApplicableIII.5.B.iv 2.4.13 Site-specific. Basis for Site Parameters ConformsAll None. 2.4.13 All RegionalGeology and Site III.2.a Conforms None. Various Not Applicable Postulated Site Parameters Not Applicable 2.4.14 Site-specific. Conforms None. 2.4.14 Not Applicable Site-specific. Not Applicable 2.4.14 Not Applicable 2.5.2 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 2.4.13, Rev 3: Accidental SRP Releases of Radioactive in Effluents Ground Liquid Waters and Surface 2.4.13, Rev 3: Accidental SRP Releases of Radioactive in Effluents Ground Liquid Waters and Surface 2.4.14, Rev 3: TechnicalSRP Specifications and Emergency Operation Requirements 2.4.14, Rev 3: TechnicalSRP Specifications and Emergency Operation Requirements 2.4.14, Rev 3: TechnicalSRP Specifications and Emergency Operation Requirements 2.4.14, Rev 3: TechnicalSRP Specifications and Emergency Operation Requirements 2.4.14, Rev 3: TechnicalSRP Specifications and Emergency Operation Requirements 2.5.1, Rev 4: Basic SRP Seismic Geologic and Information 2.5.2, Rev 4: Vibratory SRP Ground Motion 2.5.2, Rev 4: Vibratory SRP Ground Motion
Tier 2 1.9-53 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 2.5 Conforms None. 2.5.3 2.5.4 Standard (DSRS) (Continued) AC Title/Description Conformance Information Information Information AC III.2.bIII.2.cTier 1 as Included Site Parameters III.2.d TableSummary Site Parameters All Basis for Site ParametersIII.2.a ConformsIII.2.b Various None. Postulated Site ParametersIII.2.cTier 1 as Included Site Parameters ConformsIII.2.d TableSummary Site Parameters None.All Conforms Basis for Site ParametersIII.2.A None. Conforms Various Not Applicable None.III.2.B Postulated Site Parameters Site-specific. ConformsIII.2.CTier 1 as Included Site Parameters None. 2.5.2 III.2.D Conforms TableSummary Site Parameters None.All 2.5.2 Basis for Site Parameters Not Applicable Conforms Site-specific.III.2.A Not Applicable None. 2.5.3 Various Postulated Site Parameters 2.5 Conforms None. 2.5.3 Conforms None. Not Applicable 2.5.4 Not Applicable Site-specific. 2.5 2.5.4 Not Applicable 2.5.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 2.5.2, Rev 4: Vibratory SRP Ground Motion 2.5.2, Rev 4: Vibratory SRP Ground Motion 2.5.2, Rev 4: Vibratory SRP Ground Motion 2.5.3, Rev 4: Surface SRP Faulting 2.5.3, Rev 4: Surface SRP Faulting 2.5.3, Rev 4: Surface SRP Faulting 2.5.3, Rev 4: Surface SRP Faulting 2.5.3, Rev 4: Surface SRP Faulting 2.5.4, Rev 4: Stability of SRP and Subsurface Materials Foundations 2.5.4, Rev 4: Stability of SRP and Subsurface Materials Foundations 2.5.4, Rev 4: Stability of SRP and Subsurface Materials Foundations 2.5.4, Rev 4: Stability of SRP and Subsurface Materials Foundations 2.5.4, Rev 4: Stability of SRP and Subsurface Materials Foundations 2.5.5, Rev 4: Stability of SRP Slopes 2.5.5, Rev 4: Stability of SRP Slopes
Tier 2 1.9-54 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.2.1 Table 3.2-1 Table 3.2-1 gory II rather than gory II Comments Section ce criterion is applicable This acceptance criterion is applicable This which was 1.85, to RG except for reference because its guidance withdrawn in 2004 was updated and incorporated into RG 1.84. isof Table applicable A-1 butThe intent to SSC refers specific language some of the not part of the NuScale design. example,For the NuScale design does not include systems, control gas combustible emergency diesel rooms, generators, ESF or pressurizer power operated relief valves. exceptmeeting Staff Regulatory that SSC 1.29 are Guide Regulatory of C.1.i Guidance designated Seismic Cate Seismic Category I. Status Review Plan (SRP) and Design Specific Review Conforms None. 3.2.2 Conforms None. 2.5.5 Conforms None.Conforms None. 3.3.1 3.3.1 Partially Conforms This acceptan Partially Conforms Partially Conforms Partially Conforms Bounding parameters are established. 3.3.2 Standard (DSRS) (Continued) AC Title/Description Conformance Seismic Design Classification to Meet 10GDC 2; CFR Appendix A; and 100, 10 CFR S 50, Appendix Classification to Meet Group Quality and 10GDC 1 CFR 50.55a and Codes of Construction Summary for Components of Standards WaterCooled Nuclear Power Plants by NRC Quality Classification System (Page 3.2.2-12) Information for Classification Guidance Additional and Components and Systems of Application of Quality Standards Interval, and Other Site-Related Wind Interval, Parameters Pressure Equivalent Into Speed Missiles Associated AC III.2.BIII.2.CTier 1 as Included Site Parameters III.2.D TableSummary Site Parameters II.1 Basis for Site Parameters ConformsII.1 None.Table 3.2.21 Conforms None. App. A and Table A-1 2.5 2.5.5 II.1II.2 Most Severe WindII.3 DesignSpeed, Wind Recurrence II.1 ProceduresTransforming Wind for Most Severe Tornado Wind and Conforms Partially areestablished. Bounding parameters 3.3.1 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 2.5.5, Rev 4: Stability of SRP Slopes 2.5.5, Rev 4: Stability of SRP Slopes 2.5.5, Rev 4: Stability of SRP Slopes 3.2.1, Rev 2: Seismic SRP Classification 3.2.2, Rev 2: System SRP Classification Group Quality 3.2.2, Rev 2: System SRP Classification Group Quality 3.2.2, Rev 2: System SRP Classification Group Quality SRP 3.3.1, Rev. 3: SRP Wind Loadings 3.3.1, Rev. 3: SRP Wind Loadings 3.3.1, Rev. 3: SRP Wind Loadings 3.3.2: Rev. 3: SRP Tornado Loads
Tier 2 1.9-55 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.4.2 3.4.2 3.4.2 Comments Section The NuScale Certified design assumes the The NuScale Certified flood level. is above the maximum NPP design assumes the The NuScale Certified flood level. is above the maximum NPP design assumes the The NuScale Certified flood level. is above the maximum NPP Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 3.3.2 3.4.1 Conforms None.Conforms None. 3.3.2 3.3.2 Conforms Conforms Conforms None.Conforms None.Conforms None. 3.5.1 3.5.1 3.5.1 Partially Conforms Standard (DSRS) (Continued) drostatic Effects Effects drostatic Dynamic Loads AC Title/Description Conformance Requirements Demonstrating That Failure of or Component Not Structure Not Loads Will for Tornado Designed SSC to Other of the Capability Affect Functions Perform Safety with Compliance GDC 4 Conforms None. 3.4.1 Parameters and Spectrum of Parameters Tornado-Generated Missiles on Loads Into Equivalent Parameters Structures and Wave Action Consideration of Hy Consideration of Consideration of Identified Missile by Probability Analysis Missile Protection Identified Missile by Probability Analysis From Wave Action Groundwater Levels AC II.1II.2 and Classification Seismic Design II.2II.3 AcceptanceCriteria for Tornado II.4 Proceduresfor Transforming Tornado II.2II.3 Highest Flood Level Below Grade - II.1 Highest Flood Level Above Grade - II.2 Statistical Significance of an II.1 Acceptable Methods of Providing Statistical Significance of an II.1 Most Severe Highest Flood and Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.4.1, Rev. 3: Internal SRP Protection for Onsite Flood Failures Equipment 3.4.1, Rev. 3: Internal SRP Protection for Onsite Flood Failures Equipment SRP 3.3.2: Rev. 3: SRP Tornado Loads SRP 3.3.2: Rev. 3: SRP Tornado Loads 3.3.2: Rev. 3: SRP Tornado Loads SRP 3.4.2, Rev. 3: Protection of Flood from Against Structures External Sources SRP 3.4.2, Rev. 3: Protection of Flood from Against Structures External Sources 3.5.1.1, Rev. 3: Internally- SRP (Outside Missiles Generated Containment) 3.5.1.1, Rev. 3: Internally- SRP (Outside Missiles Generated Containment) 3.5.1.2, Rev. 3: Internally SRP (Inside Missiles Generated Containment) SRP 3.4.2, Rev. 3: Protection of Flood from Against Structures External Sources
Tier 2 1.9-56 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Comments Section Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Status Review Plan (SRP) and Design Specific Review Conforms None. 3.5.1 Not Applicable Section I and DSRS 3.5.1.3, DSRS 10.2.3, Per Not Applicable Section I and DSRS 3.5.1.3, DSRS 10.2.3, Per Not Applicable Section I and DSRS 3.5.1.3, DSRS 10.2.3, Per Not Applicable Section I and DSRS 3.5.1.3, DSRS 10.2.3, Per Standard (DSRS) (Continued) Probability Tables AC Title/Description Conformance Missile Protection From Turbine Missiles Probability ObtainingProgram for Applicants Manufacturers without Turbine From for Procedures NRC-Approved Calculating Missile Generation Probabilities From Turbine Manufacturers Table 3.5.1.3-1) (Including AC II.2II.1 Acceptable Methods of Providing Damage Unacceptable of Probability II.2 Turbine Missile GenerationII.3 Not Applicable AcceptablyLow Missile Generation Section I and DSRS 3.5.1.3, DSRS 10.2.3, Per II.4 Missile Generation II.5 Inservice Inspection and Test II.6 Protective Barriers Conforms None. 3.5.1.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.5.1.2, Rev. 3: Internally 3.5.1.2, Rev. 3: Internally SRP (Inside Missiles Generated Containment) DSRS 3.5.1.3, Rev. 0: Turbine Missiles DSRS 3.5.1.3, Rev. 0: Turbine Missiles DSRS 3.5.1.3, Rev. 0: Turbine Missiles DSRS 3.5.1.3, Rev. 0: Turbine Missiles DSRS 3.5.1.3, Rev. 0: Turbine Missiles DSRS 3.5.1.3, Rev. 0: Turbine Missiles
Tier 2 1.9-57 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.5.1.4 Table 2.0-1 Table 2.0-1 Table 2.0-1 Table 2.0-1 Not Applicable Not Applicable Not Applicable Not Applicable also includes RG 1.221 also includes RG 1.221 Comments Section for Design Basis Hurricane-Generated Design for Missiles. The NuScale design assumes no proximity The NuScale design assumes missiles. no proximity The NuScale design assumes missiles. The NuScale design assumesno aircraft hazard missiles. The NuScale design assumesno aircraft hazard missiles. The NuScale design assumesno aircraft hazard missiles. The NuScale design assumesno aircraft hazard missiles. The NuScale design assumesno aircraft hazard missiles. acceptance criterion specifiesThis provisions when using composite or multi- not use element barriers. NuScale does composite or multi-element barriers. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 3.5.2 3.5.3 Conforms The NuScale design Conforms Conforms None.Conforms None. 3.5.1.4 3.5.1.4 Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Concrete sections Composite Compliancewith 10 CFR 100Compliance with GDC 4 Not Applicable Not Applicable CapabilitySSC to Withstand the of Generated Externally of Effects Missiles Missile Spectrum Information Identified Missile by Probability by Natural Missiles Generated Phenomena AC II.1.AII.1.B Local For Prediction - Damage II.1.C SteelPrediction - Local Damage For Local For Prediction - Damage Conforms None. 3.5.3 II (no number) II (no II.1 Design Basis Tornado-Generated II.1 II.2 II.1 and II.2 VariousIII.8.B.1III.8.B.2 Postulated Site ParametersIII.8.B.3Tier 1 as Included Site Parameters III.8.B.4 TableSummary Site Parameters Conforms Basis for Site Parameters Applicable Not Conforms Conforms II.2II.3 Statistical Significance of an Design Basis Appropriate Identifying Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.5.3, Rev. 3: Barrier SRP Design Procedures 3.5.3, Rev. 3: Barrier SRP Design Procedures 3.5.3, Rev. 3: Barrier SRP Design Procedures SRP 3.5.1.4, Rev. 4: Missiles SRP Winds Generated by Extreme 3.5.1.5, Rev 4: Site SRP (Except Proximity Missiles Aircraft) 3.5.1.5, Rev 4: Site SRP (Except Proximity Missiles Aircraft) 3.5.1.6, Rev 4: Aircraft SRP Hazards 3.5.1.6, Rev 4: Aircraft SRP Hazards 3.5.1.6, Rev 4: Aircraft SRP Hazards 3.5.1.6, Rev 4: Aircraft SRP Hazards 3.5.1.6, Rev 4: Aircraft SRP Hazards 3.5.2, Rev 3: Structures, SRP ComponentsSystems, and to be Protected From Externally- Generated Missiles SRP 3.5.1.4, Rev. 4: Missiles SRP Winds Generated by Extreme 3.5.1.4, Rev. 4: Missiles SRP Winds Generated by Extreme
Tier 2 1.9-58 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.5.3 Comments Section This acceptance criterion is applicable This to subtier ANSI/AISC except for reference 2 (2004). with Supplement N690-1994 this of version NuScale uses the 2012 standard. Status Review Plan (SRP) and Design Specific Review Conforms None. 3.6.1 Conforms None.Conforms None. 3.6.1 Conforms None. 3.6.1 3.6.2 ediction Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Separation of High and Moderate Moderate High and Separation of Energy Fluid Systems From Essential Systems/Components Systems AreEnclosed Separation Nor Protective Enclosures Are Practical Inside Containment AC II.2 Overall Damage For Pr II.1 II.2 Fluid Energy Moderate High and II.3 Cases Where Neither Physical II.4 Design FeaturesII.5 Failures Postulated of Effects II.1 Conforms Postulated Pipe RuptureLocations Conforms None. None. 3.6.1 3.6.1 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.5.3, Rev. 3: Barrier SRP Design Procedures Design 3.6.1, Rev 3: Plant SRP Against for Protection Piping Failures in Postulated Fluid Systems Outside Containment Design 3.6.1, Rev 3: Plant SRP Against for Protection Piping Failures in Postulated Fluid Systems Outside Containment Design 3.6.1, Rev 3: Plant SRP Against for Protection Piping Failures in Postulated Fluid Systems Outside Containment Design 3.6.1, Rev 3: Plant SRP Against for Protection Piping Failures in Postulated Fluid Systems Outside Containment Design 3.6.1, Rev 3: Plant SRP Against for Protection Piping Failures in Postulated Fluid Systems Outside Containment 3.6.2, Rev 2: SRP Determination of Rupture Dynamic Locations and the with Associated Effects of Piping Rupture Postulated
Tier 2 1.9-59 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None. 3.6.2 Conforms None.Conforms None. 3.6.2 3.6.2 Standard (DSRS) (Continued) AC Title/Description Conformance Compliance with Compliance GDC 4 Conforms None. 3.6.3 Outside Containment Impingement Forces Components on Mechanical Effects and Supports AC II.2 Postulated Pipe RuptureLocations II.3 Methods of AnalysisIII.1 Pipe Break CriteriaIII.2 Conforms Dynamic Effects None.III.3 ConformsModeling Jet for Assumptions None.III.4 Conforms Dynamic Break of Pipe Analyses II.1 None.II.2II.1 3.6.2 Low Probability of Pipe Rupture Design Ground Motion 3.6.2 Conforms None. Conforms 3.6.2 None. 3.6.3 3.7.1 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.6.2, Rev 2: SRP Determination of Rupture Dynamic Locations and the with Associated Effects of Piping Rupture Postulated 3.6.2, Rev 2: SRP Determination of Rupture Dynamic Locations and the with Associated Effects of Piping Rupture Postulated 3.6.2, Rev 2: SRP Determination of Rupture Dynamic Locations and the with Associated Effects of Piping Rupture Postulated 3.6.2, Rev 2: SRP Determination of Rupture Dynamic Locations and the with Associated Effects of Piping Rupture Postulated 3.6.2, Rev 2: SRP Determination of Rupture Dynamic Locations and the with Associated Effects of Piping Rupture Postulated 3.6.2, Rev 2: SRP Determination of Rupture Dynamic Locations and the with Associated Effects of Piping Rupture Postulated 3.6.3, Rev 1: Leak-Before- SRP Break Evaluation Procedures 3.6.3, Rev 1: Leak-Before- SRP Break Evaluation Procedures DSRS 3.7.1, Rev. 0: Seismic DSRS Design Parameters
Tier 2 1.9-60 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.7.2 Not Applicable Comments Section NuScale does not perform both time history analysis and responsespectrum analysis in its analysis of structures. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None.Conforms None.Conforms 3.7.1 None.Conforms None.Conforms None.Conforms 3.7.2 None. 3.7.2 3.7.2 3.7.2 3.7.2 3.7.2 3.7.2 Conforms None.Conforms None. 3.7.1 3.7.1 Standard (DSRS) (Continued) ter Variations on ter Variations AC Title/Description Conformance Modeling Response Spectra Motion Floor Response Spectra Factors Torsional Effects Review ConsiderationsDC and for COL Applications Interaction of Non-Seismic Category I I Category with Seismic Structures SSCs Values I Category Structures AC II.1II.2II.3 Seismic Analysis MethodsII.4 Natural Frequencies and ResponsesII.5 Procedures Used for Analytical ConformsII.6 Conforms Soil-Structure Interaction None.II.7 None.In-Structure Development of II.8 Three Components of Design Ground Responses Modal of Combination ConformsII.9 None.II.10 ConformsII.11 of Parame Effects None. Static Vertical Equivalent of Use II.12 for to Account Used Methods 3.7.2 Responsesof Comparison 3.7.2 Applicable Not 3.7.2 3.7.2 II.2II.3II.4 Damping Percentage of Critical Seismic Supporting Media for II.13Damping Analysis Procedure for Conforms None. 3.7.1 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.2, Rev 0: Seismic DSRS System Analysis DSRS 3.7.1, Rev. 0: Seismic DSRS Design Parameters DSRS 3.7.1, Rev. 0: Seismic DSRS Design Parameters 3.7.1, Rev. 0: Seismic DSRS Design Parameters DSRS 3.7.2, Rev 0: Seismic DSRS System Analysis
Tier 2 1.9-61 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section The NuScale design does not usedams. Not Applicable Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None.Conforms None. 3.7.3 3.7.3 3.7.3 3.7.3 Conforms None.Conforms None.Conforms None.Conforms None. 3.7.2 3.7.3 3.7.3 3.7.3 Conforms None. 3.7.3 Not Applicable Frequencies Conforms None. 3.7.3 Standard (DSRS) (Continued) AC Title/Description Conformance Components with Distinct Inputs with Distinct Components Factors Conduits, and Tunnels I Concrete Dams Seismic Category Interaction of Non-Seismic Category I Subsystems with Seismic Category I SSCs Moments and Sliding Forces, Moments and and to Soil Pressures Structure Frictional Forces for Seismic Category I Structures Earthquake Cycles Modeling Motion Above-Ground Tanks AC II.9II.10II.11 Multiply-Supported Equipment and Static Vertical Equivalent of Use II.12 TorsionalMasses Effects of Eccentric II.13 Buried Piping, I Seismic Category Conforms Methods for Seismic Analysis of None. 3.7.3 II.14 Determination of Overturning II.1II.2II.3 Seismic Analysis MethodsII.4 Determination of Number II.5 Procedures Used for Analytical II.6 Conformsfor Selection of Basis II.7 None. Analysis ProcedureDamping for II.8 Three Components of Design Ground Conforms Responses Modal of Combination None. Conforms None. 3.7.3 3.7.3 3.7.3 II.14 Methods for Seismic Analysis of Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis DSRS 3.7.2, Rev 0: Seismic DSRS System Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis DSRS 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis DSRS 3.7.3, Rev. 0: Seismic DSRS Subsystem Analysis
Tier 2 1.9-62 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.7.4 Not Applicable Not Applicable Not Applicable Comments Section ee RG 1.166 in TableRG 1.166 ee 1.9-2. Not Applicable There is a COL item to comply. Locations are to comply. Locations COL item a There is identified in conformanceRG with 1.12, however seismic instrumentation cannot be placed inside containment. expectationIdentified as anCOL for applicants. The NuScale design does not have a concrete containment. not have does The NuScale containment internal structures. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 3.8.2 Conforms None. 3.8.2 3.8.2 3.8.4 Not Applicable inations Conforms None. 3.8.2 Standard (DSRS) (Continued) AC Title/Description Conformance Comparison with Comparison RG 1.12 with Comparison RG 1.166 Partially Conforms Comparisonthe requirements of with 10 CFR (ALARA) 20.1101 Not Applicable S Specifications Construction Techniques Requirements Specifications AC II.1 II.2 II.3 All Various Not Applicable All Various Not Applicable II.1II.2II.3the Containment Description of II.4 and Standards, Codes, Applicable II.5 Conforms Loads and Loading Comb II.6 None. Design and Analysis ProceduresII.7 Acceptance Criteria Structural Conforms Materials, Quality Control, and Special None. TestingSurveillance and Inservice ConformsII.1 None.II.2II.3the Structures Description of II.3.A and Standards, Codes, Applicable 3.8.2 Loads and Load Combinations Conforms Concrete Structures None. Conforms 3.8.2 None. 3.8.2 Conforms None. 3.8.4 3.8.4 3.8.4 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.7.4, Rev 2: Seismic SRP Instrumentation 3.7.4, Rev 2: Seismic SRP Instrumentation 3.7.4, Rev 2: Seismic SRP Instrumentation 3.8.1, Rev 4: Concrete SRP Containment and Rev 4: Concrete SRP 3.8.3, of Steel Internal Structures Steel or Concrete Containments DSRS 3.8.2, Rev. 0: Steel DSRS Containment 3.8.2, Rev. 0: Steel DSRS Containment 3.8.2, Rev. 0: Steel DSRS Containment 3.8.2, Rev. 0: Steel DSRS Containment 3.8.2, Rev. 0: Steel DSRS Containment 3.8.2, Rev. 0: Steel DSRS Containment 3.8.2, Rev. 0: Steel DSRS Containment 3.8.4, Rev. 0, Other DSRS I Structures Seismic Category 3.8.4, Rev. 0, Other DSRS I Structures Seismic Category 3.8.4, Rev. 0, Other DSRS I Structures Seismic Category 3.8.4, Rev. 0, Other DSRS I Structures Seismic Category
Tier 2 1.9-63 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Comments Section ress Analysis Method is not ress Analysis used. Masonry walls Masonry are not used in the NuScale design. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 3.8.4 3.8.4 Conforms None. 3.9.1 Conforms None.Conforms None.Conforms None. 3.8.5 3.8.5 3.8.5 Not Applicable Experimental St Standard (DSRS) (Continued) AC Title/Description Conformance Construction Techniques Requirements Analyses Dynamic and Static in Lieu of Analytical Methods Methods Specifications Construction Techniques Requirements AC II.3.BII.4 Steel StructuresII.5II.6 Design and Analysis ProceduresII.7 Acceptance Criteria Structural II.8 Conforms Materials, Quality Control, and Special None. TestingSurveillance and Inservice Conforms Conforms Masonry Walls None. None. Not Applicable II.1 3.8.4 II.2 3.8.4 3.8.4 SpecificationTransients of II.3 in Used be to Programs Computer Conforms Analysis Stress Experimental of Use None. 3.9.1 II.1II.2II.3 Descriptionthe Foundation of II.4 and Standards, Codes, Applicable II.5 Loads and Load Combinations ConformsII.6 Design and Analysis Procedures None.II.7 Acceptance Criteria Structural Conforms Conforms Materials, Quality Control, and Special None. None. TestingSurveillance and Inservice Conforms None. 3.8.5 3.8.5 3.8.5 3.8.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 3.8.4, Rev. 0, Other DSRS I Structures Seismic Category 3.9.1, Rev 3: Special SRP TopicsMechanical for Components 3.9.1, Rev 3: Special SRP TopicsMechanical for Components 3.9.1, Rev 3: Special SRP TopicsMechanical for Components DSRS 3.8.4, Rev. 0, Other DSRS I Structures Seismic Category 3.8.4, Rev. 0, Other DSRS I Structures Seismic Category 3.8.4, Rev. 0, Other DSRS I Structures Seismic Category 3.8.4, Rev. 0, Other DSRS I Structures Seismic Category 3.8.4, Rev. 0, Other DSRS I Structures Seismic Category DSRS 3.8.5, Rev. 0: DSRS Foundations 3.8.5, Rev. 0: DSRS Foundations 3.8.5, Rev. 0: DSRS Foundations 3.8.5, Rev. 0: DSRS Foundations 3.8.5, Rev. 0: DSRS Foundations 3.8.5, Rev. 0: DSRS Foundations 3.8.5, Rev. 0: DSRS Foundations
Tier 2 1.9-64 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.9.2 Comments Section This acceptance criterion is applicable This except for aspects related to test performance and associated corrective actions (as required), which are the applicantresponsibilitythe COL of referencing the certified design. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms 3.9.1 None.Conforms None.Conforms None.Conforms 3.9.2 None. 3.9.2 3.9.2 3.9.2 3.9.2 Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Compliance with Compliance GDC 2 Conforms None. 3.9.2 Specified for Codeand Class 1 Core Support Components Testing Dynamic Effects Reactor InternalsVibrations for of Prototype Plants Program Test Reactor CoolantReactor Internals and Piping Reactor Internals Applications AC II.4II.1 When Service Level D Limits are Vibration, Thermal Expansion, and II.2 II.3II.4 to Predict Solutions Analytical II.5 Preoperational Vibration and Stress II.6 of Design Adequacy Structural II.7 Correlation of Tests and Analyses of Test for New Specifications Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.9.1, Rev 3: Special SRP TopicsMechanical for Components 3.9.2, Rev 3: Dynamic SRP of Testing and Analysis Systems, Structures, and Components 3.9.2, Rev 3: Dynamic SRP of Testing and Analysis Systems, Structures, and Components 3.9.2, Rev 3: Dynamic SRP of Testing and Analysis Systems, Structures, and Components 3.9.2, Rev 3: Dynamic SRP of Testing and Analysis Systems, Structures, and Components 3.9.2, Rev 3: Dynamic SRP of Testing and Analysis Systems, Structures, and Components 3.9.2, Rev 3: Dynamic SRP of Testing and Analysis Systems, Structures, and Components 3.9.2, Rev 3: Dynamic SRP of Testing and Analysis Systems, Structures, and Components
Tier 2 1.9-65 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.9.4 3.9.5 3.9.5 Not Applicable t is the issue of this of t is the issue Comments Section NRC Bulletin 88-11 appliesPWR designs to that incorporatea pressurizer separatefrom the reactor pressurevessel, with a surge line connecting theNuScaletwo. In the design, the pressurizer (i.e., is located is integral reactorwithin) to the pressure vessel: there is no pressurizersurge line within which thermal stratification (tha occur. would bulletin) acceptance criterion is applicable This (seismic design per RG 1.29) but contains a wordingtypographical is error. The because it mixes an SRP section confusing reference with a RG. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 3.9.4 3.9.3 Conforms None.Conforms None. 3.9.3 3.9.3 Conforms None.Conforms None. 3.9.4 3.9.3 Standard (DSRS) (Continued) AC Title/Description Conformance LoadSets for Design Combination and Service Conditions Defined in NB-3113 Section III, ASME Code Combinations, andLoads, Loading to Limits for Portions Constructed ASME CodeSection NG Operating Transients, and Stress Limits Relief Devices Construction Support Structures AC II.1II.2 Loading Combinations, System II.3 Pressure of and Installation Design Component SupportsII.1 Not Applicable Adequacy of Descriptive InformationII.2 Conforms II.3 Codes and Standards for II.4II.1 Operability Assurance ProgramII.2 Conforms Core of and Construction Design None. 3.9.4 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.9.3, Rev 3: ASME Code 3.9.3, Rev 3: ASME SRP 1, 2, and 3 Components Class and Component Supports, Structures Core Support and Code 3.9.3, Rev 3: ASME SRP 1, 2, and 3 Components Class and Component Supports, Structures Core Support and Code 3.9.3, Rev 3: ASME SRP 1, 2, and 3 Components Class and Component Supports, Structures Core Support and 3.9.4, Rev 3: Control Rod SRP Drive Systems 3.9.4, Rev 3: Control Rod SRP Drive Systems 3.9.4, Rev 3: Control Rod SRP Drive Systems 3.9.4, Rev 3: Control Rod SRP Drive Systems 3.9.5, Rev 3: Reactor SRP Pressure Vessel Internals 3.9.5, Rev 3: Reactor SRP Pressure Vessel Internals
Tier 2 1.9-66 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.9.5 3.9.5 3.9.6 ve, this guidance is ve, this guidance provides Comments Section The intent of subtier NUREG-0609 is subtier NUREG-0609 of The intent applicablebut the language refers toa conditions and SSC LWR of type different relevant to the NuScale design. not Specifically, this methodology for evaluation of loading transients and structural components, including containment subcompartment analysis, when a double-ended guillotine loopbreak of reactor coolantoccurspiping the reactor vessel inlet. The NuScale at containment vessel design does not have subcompartments. In addition, the NuScale design does not have reactor coolant loops. Notwithstanding the abo loadingapplicable to the evaluation of transients and structural components for postulated breaksand of chemical volume control system (CVCS)piping piping and at the reactor vent valves. acceptanceThis criterion (including Appendix A) is applicable except for aspects that are BWR-specific. acceptance criterion is applicable This functional to for aspects related except design,qualification, of safety- and testing related pumps. Safety-related pumps are design. used in the NuScale not Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 3.9.2 3.9.5 Partially Conforms Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance and Analyses for Design of Reactor for Design of Reactor Analyses and Than Core Support Internals Other Structures Internals Asymmetric Accommodate BlowdownLoads From Postulated Pipe Ruptures Acoustic Resonances (Including A) Appendix of Pumps, Valves, and Dynamic Restraints AC II.3II.4 Design Criteria, Loading Conditions, II.5 Deformation Limits for Reactor Internals to of Reactor Design II.6II.1 and Vibration of Flow-Induced Effects Functional Design and Qualification Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.9.5, Rev 3: Reactor 3.9.5, Rev 3: Reactor SRP Pressure Vessel Internals 3.9.5, Rev 3: Reactor SRP Pressure Vessel Internals 3.9.5, Rev 3: Reactor SRP Pressure Vessel Internals 3.9.5, Rev 3: Reactor SRP Pressure Vessel Internals 3.9.6, Rev 3: Functional SRP Design, Qualification, and Inservice Testing Programs and Valves, for Pumps, Dynamic Restraints
Tier 2 1.9-67 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.9.6 3.9.6 3.9.6 Not Applicable Not Applicable Comments Section ction 3.9.6.3.2 for valve testing for valve ction 3.9.6.3.2 pumps or dynamic restraints that perform a OM ASME in the identified specific function Code Subsection ISTA-1100. and alternative authorizations to the code. This acceptance criterion is applicable This except for aspects related to inservice testing of safety-related pumps. Safety- NuScalerelated pumps are not used in the design.The only pumps thatwithin fall the the NuScale design inof this criterionscope pumps. These pumps are are the CVCS containASME Class reactor they III because during normalcoolant operation, but they function.serve no safety acceptance criterion is relatedThis to including operational activities, testing, implementation of pre-service inservice testing and inspection, motor- valvethat are testing programs,operated the responsibilityCOL of the applicant referencing the certified design. and Section 3.9.6.6 for augmented valves for augmented Section 3.9.6.6 and testing program. Status Review Plan (SRP) and Design Specific Review Conforms requests for relief Section 3.9.6.5 Refer to Not Applicable The NuScale Power Plant does not have lves Partially Conforms Refer to Se for PumpsConforms Partially Standard (DSRS) (Continued) AC Title/Description Conformance Dynamic Restraints Alternatives AC II.2 Program Inservice Testing II.3for Va Program Inservice Testing II.4for Program Inservice Testing II.5 Relief Requests and Proposed II.6 Operational Programs Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.9.6, Rev 3: Functional SRP Design, Qualification, and Inservice Testing Programs and Valves, for Pumps, Dynamic Restraints 3.9.6, Rev 3: Functional SRP Design, Qualification, and Inservice Testing Programs and Valves, for Pumps, Dynamic Restraints 3.9.6, Rev 3: Functional SRP Design, Qualification, and Inservice Testing Programs and Valves, for Pumps, Dynamic Restraints 3.9.6, Rev 3: Functional SRP Design, Qualification, and Inservice Testing Programs and Valves, for Pumps, Dynamic Restraints 3.9.6, Rev 3: Functional SRP Design, Qualification, and Inservice Testing Programs and Valves, for Pumps, Dynamic Restraints
Tier 2 1.9-68 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.10 Not Applicable Not Applicable Not Applicable Comments Section sed seismic qualification is Developmentrisk-implementation of a and informed, performance-based inservice testing program is the responsibilityCOL of applicants that reference the NuScale certified design and that elect to implement such a program. Developmentrisk-implementation of a and for informed, inservice inspectionprogram piping is the responsibilityof COL applicants that reference the NuScale certified design, and that elect to implement such a program. Records indicates that a The NuScale design COL program is required and includes a item to maintain one. not used. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms 3.10 None.Conforms None. 3.10 3.10 Partially Conforms Partially in TableRG 1.97 See 1.9-2 3.11 n Not Applicable Experience ba Standard (DSRS) (Continued) AC Title/Description Conformance Testing of Instrumentation Described in RG 1.97 and Associated Supports are Part of the Reactor Coolant Pressure Boundary Program AC All VariousAll VariousII.1 Not Applicable II.2 QualificationElectrical of Equipment II.3 Not Applicable II.4 Experience-Based Qualificatio II.5 RecordsII.6 for Valves that Qualification Program DocumentationQualification of Conforms Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 3.9.7, Rev 0: Risk- SRP Informed Inservice Testing 3.9.8, Rev 0: Standard SRP Review Plan for the Review of Risk-Informed Inservice Inspection of Piping 3.10, Rev 3: Seismic and SRP Dynamic Qualification of Mechanical and Electrical Equipment 3.10, Rev 3: Seismic and SRP Dynamic Qualification of Mechanical and Electrical Equipment 3.10, Rev 3: Seismic and SRP Dynamic Qualification of Mechanical and Electrical Equipment 3.10, Rev 3: Seismic and SRP Dynamic Qualification of Mechanical and Electrical Equipment 3.10, Rev 3: Seismic and SRP Dynamic Qualification of Mechanical and Electrical Equipment 3.10, Rev 3: Seismic and SRP Dynamic Qualification of Mechanical and Electrical Equipment
Tier 2 1.9-69 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.11.2 Comments Section IEEE is applicable. See RG 317-1983 1.63 in Tableentry the to with respect 1.9-2 aspects of RGother 1.63. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms The portionof the guidance that endorses Conforms None. 3.11 3.11.2 Conforms None.Conforms None. 3.11 3.11 Partially ConformsPartially in TableRG 1.89 See 1.9-2. 3.11 Partially Conforms Partially in TableRG 1.89 See 1.9-2. 3.11.2 Partially Conforms Partially in TableRG 1.97 See 1.9-2. 3.11.2 Standard (DSRS) (Continued) ication Related to AC Title/Description Conformance Application of RG 1.89 for 1.89 Application of RG Environmental Qualification Program per 10 CFR 50.49 Application of Clarif IEEE Std. 323 Criteria Std. 323 IEEE Environmental Design and ElectricalQualification of Penetration Assemblies Environmental Design and Electric Valve Class 1E Qualification of Operators Environmental Qualification of Electrical Equipment Importantto Safety Environmental Design and PostAccident Qualification of Monitoring Equipment Environmental designand qualification of computer-specific requirements Environmental designand qualification of power, andinstrumentation, control portions the safety systems of AC II.3II.4 Application of RG 1.63 for II.5 Application of RG 1.73 for Application of RG 1.89 for II.1 II.2 II.6 Application of RG 1.97 for II.7II.8 Application of RG 1.152 for Application of RG 1.153 for Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment
Tier 2 1.9-70 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Comments Section because it is associatedis with an external/ because it natural event. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 3.11 3.11.2 3.11.2 Not Applicable in TableRG 1.158 See 1.9-2. Not Applicable Not Applicable Lightning protection is not applicable to EQ Partially Conforms Partially in TableRG 1.180 See 1.9-2. 3.11.2 Partially Conforms Partially in TableRG 1.100 See 1.9-2. 3.11 Partially Conforms Partially in TableRG 1.183 See 1.9-2. 3.11.2 Standard (DSRS) (Continued) AC Title/Description Conformance Application of RG 1.180 for Application of RG 1.180 Electromagnetic and Radio- Frequency Interference in Safety Related I&C Equipment Environmental designand qualification of safety-related computer-based systems in mild I&C environments Environmental Qualification of Class CablesSplices and1E Electric Field Environmental Qualification of Class 1E Connection Assemblies Environmental Qualification of Class Batteries Storage 1E Lead ElectricalQualification of and Active andMechanical Equipment Functional Qualification of Active for Nuclear Mechanical Equipment Power Plants Source Term Used in Environmental Design and Qualification of to Safety Important Equipment Environmental designand qualification of the lightning system protection AC II.9 Application of RG 1.209 for II.10 for Application of RG 1.211 II.11 for Application of RG 1.156 II.12 for Application of RG 1.158 II.13 II.15 for Seismic Application of RG 1.100 II.14 for Accident Application of RG 1.183 II.16 for Application of RG 1.204 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment
Tier 2 1.9-71 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable t item. Not Applicable Comments Section The programs are described and maintained by the COL applicant. is a COL applican This Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 3.11 3.11 Conforms None.Conforms None. 3.11 3.11 Conforms None. 3.11 Not Applicable Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance to provideAssurance assurance of Environmental Design and Qualificationof StatusEquipment in Harsh Environments Mild and Design/Purchase Specifications of Perform Under Equipment to Applicable Environmental Conditions Applicable documentation for Environmental Design and Safety-RelatedQualification of Mechanical, Electrical, and I&C Equipment for All Important to Safety Equipment Safety Important to for All to Safety- Essential Equipment Functions Related Implementation Features Systems Engineered Safety AC II.21 II.22 Maintenance/surveillance programs II.17 Conditions of Environmental Effects II.18 Suitabilityand of Materials, Parts, II.19 QualificationNonmetallic Parts of II.20 Conforms None. 3.11 II.23 Operational Program II.24Organic Exposure of Components on Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment
Tier 2 1.9-72 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable t item. Not Applicable Comments Section This is a COL applican This This guidance is applicable only to BWR plants. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 3.13.1 3.13.2 Not Applicable iteria Conforms None. 3.12.5 Standard (DSRS) (Continued) AC Title/Description Conformance Specifications 1) Requirements (Including Table 3.13- 2) AC II.25 Design and Procurement II.A PipingMethods Analysis II.B PipingTechniques Modeling II.C Conforms None. PipingCr Stress Analysis ConformsII.D None. Piping Support DesignII.1II.2 Design Aspects(Including Table 3.13- All ConformsInspection Preservice and Inservice 3.12.3 None. 3.12.4 3.12.6 Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 3.11, Rev. 0: DSRS Environmental Qualification Electricalof Mechanical and Equipment Code 3.12, Rev 1: ASME SRP 1, 2, and 3 Piping Class Components Systems, Piping andAssociated Their Supports Code 3.12, Rev 1: ASME SRP 1, 2, and 3 Piping Class Components Systems, Piping andAssociated Their Supports Code 3.12, Rev 1: ASME SRP 1, 2, and 3 Piping Class Components Systems, Piping andAssociated Their Supports Code 3.12, Rev 1: ASME SRP 1, 2, and 3 Piping Class Components Systems, Piping andAssociated Their Supports 3.13, Rev. 0: Threaded SRP FastenersClass - ASME Code 3 1, 2, and 3.13, Rev. 0: Threaded SRP FastenersClass - ASME Code 3 1, 2, and 2: Classification Rev BTP 3-1, of Main Steam Components OtherReactor Than the Coolant Pressure Boundary Plants for BWR
Tier 2 1.9-73 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.1 15.2 15.5 15.6 15.1 15.2 15.5 15.6 15.1 15.2 15.5 15.6 Not Applicable Comments Section This guidance is applicable only to BWR plants. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms 3.6 None. 3.6 stem Piping Conforms None. 3.6 Standard (DSRS) (Continued) AC Title/Description Conformance Piping Failures Piping Piping Fluid System AC AllB.1B.2 Plant ArrangementB.3 Design FeaturesB.4 of Postulated Analyses and Effects Conforms Not Applicable None.B.A Implementation Conforms None. High-Energy Fluid System PipingB.B Conforms Moderate-Energy Fluid Sy ConformsB.C None. None. Typeof Breaks and Leakage Cracks in 3.6 3.6 3.6 3.6 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: BTP 3-2, Rev 2: Classification Rev BTP 3-2, Steam and BWR/6 Main of Feedwater Components OtherReactor Than the Coolant Pressure Boundary 3: Protection Rev BTP 3-3, Piping Postulated Against Failures in Fluid Systems Outside Containment 3: Protection Rev BTP 3-3, Piping Postulated Against Failures in Fluid Systems Outside Containment 3: Protection Rev BTP 3-3, Piping Postulated Against Failures in Fluid Systems Outside Containment 3: Protection Rev BTP 3-3, Piping Postulated Against Failures in Fluid Systems Outside Containment 2: Postulated Rev BTP 3-4, Locations in Fluid Rupture System Piping Inside and Outside Containment 2: Postulated Rev BTP 3-4, Locations in Fluid Rupture System Piping Inside and Outside Containment 2: Postulated Rev BTP 3-4, Locations in Fluid Rupture System Piping Inside and Outside Containment
Tier 2 1.9-74 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 4.2.3 4.2.3 4.2.4 4.2.4 4.4.2 15.0.0 Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None. 4.2.1 Conforms None.Conforms None. 4.2.1 4.2.1 Conforms None.Conforms None.Conforms None. 4.3.1 4.3.2 4.4.1 Conforms None. 4.4.7 is Codes Conforms None. 4.4.4 Standard (DSRS) (Continued) AC Title/Description Conformance Plans Structural Response to Externally Externally Response to Structural Applied Forces Initiated the Reactivity for Guidance Accidents Power Distributions Power Worth Thermal Margin Hydraulic Instabilities AC II.1II.1.AII.1.B Design Bases Damage System Fuel Fuel Rod Failure Conforms Conforms None. None. Conforms None. 4.2.1 4.2.1 4.2.1 II.1.CII.2 Fuel CoolabilityII.3Design Drawings Description and Design Evaluation Conforms None. Conforms None. Conforms None. 4.2.2 4.2.1 4.2.1 II.4 Testing, Inspection, and Surveillance App AApp B Evaluation of Fuel Assembly Interim Acceptanceand Criteria II.1II.2II.3 Design Limits for Power Densities and II.4 Reactivity CoefficientsII.1 Control Rod Patterns and Reactivity Data and Methods Analytical Core Design, and Fuel Design Limits, Conforms Conforms None. None. 4.3.2 4.3.3 II.2II.3 Subchannel Hydraulic Analys Thermal- and Oscillations Core Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 4.2, Rev 3: Fuel System 3: Fuel System 4.2, Rev SRP Design SRP 4.2, Rev 3: Fuel System 3: Fuel System 4.2, Rev SRP Design 3: Fuel System 4.2, Rev SRP Design SRP 4.2, Rev 3: Fuel System 3: Fuel System 4.2, Rev SRP Design 3: Fuel System 4.2, Rev SRP Design 3: Fuel System 4.2, Rev SRP Design SRP 4.2, Rev 3: Fuel System 3: Fuel System 4.2, Rev SRP Design SRP 4.2, Rev 3: Fuel System 3: Fuel System 4.2, Rev SRP Design 3: Fuel System 4.2, Rev SRP Design SRP 4.3, Rev 0: Nuclear Design SRP 4.3, Rev 0: Nuclear Design SRP 4.3, Rev 0: Nuclear Design SRP 4.3, Rev 0: Nuclear Design 4.4, Rev 0: Thermal and DSRS Hydraulic Design DSRS 4.4, Rev 0: Thermal and DSRS Hydraulic Design 4.4, Rev 0: Thermal and DSRS Hydraulic Design
Tier 2 1.9-75 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 16.1 4.5.1 4.5.1 4.5.1 4.4.6 4.4.6 4.4.4 4.4.6 Not Applicable Comments Section Diversepreventsignals an ATWS from RTS This prevents flowoccurring. instabilities from occurring, so this AC is not applicable based on thecurrent ATWS approach. 2004. Guidance was withdrawn in RG 1.85 was updated and incorporated into RG 1.84. on ANSI/ASME is based QAPD The NuScale as addenda, NQA-1a-2009 with NQA-1-2008 Rev. 4. endorsed by RG 1.28, on ANSI/ASME is based QAPD The NuScale as addenda, NQA-1a-2009 with NQA-1-2008 Rev. 4. endorsed by RG 1.28, damage from loose parts and the need for damagethe need from loose parts and loose parts monitoring system. e flow Low precludes in primary systems Status Review Plan (SRP) and Design Specific Review Conforms Conforms None.Conforms None. 4.4.5 4.4.2 Conforms None. 4.4.6 Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Components Programs Process Monitoring Detection and Recovery from Detection and Cooling Core Inadequate Scram without Transient Anticipated Event AC II.4II.5 RPV Fluid Flow Calculations Technical Specifications Conforms None. Conforms None.II.1II.2 Materials SpecificationsII.3 Stainless Austenitic Steel II.4 4.4.4 Other Materials 4.4.3 Conforms Cleanliness Control Cleaning and Conforms Conforms None. 4.5.1 II.1II.2 Materials Controls on Welding Conforms Conforms None. None. 4.5.2 4.5.2 II.6II.7 Test and Initial Preoperational II.8Monitoring System Loose Parts and Calculations Heat Flux Critical Departur II.9II.10 Instrumentationand Procedures for Core Stability Performance During Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 4.4, Rev 0: Thermal and DSRS Hydraulic Design 4.5.1, Rev 3: Control Rod SRP Drive Structural Materials 4.5.1, Rev 3: Control Rod SRP Drive Structural Materials 4.5.1, Rev 3: Control Rod SRP Drive Structural Materials 4.5.1, Rev 3: Control Rod SRP Drive Structural Materials DSRS 4.4, Rev 0: Thermal and DSRS Hydraulic Design SRP 4.5.2, Rev 3: Reactor 4.5.2, Rev 3: Reactor SRP and Core Support Internal Materials Structure 4.5.2, Rev 3: Reactor SRP and Core Support Internal Materials Structure DSRS 4.4, Rev 0: Thermal and DSRS Hydraulic Design 4.4, Rev 0: Thermal and DSRS Hydraulic Design 4.4, Rev 0: Thermal and DSRS Hydraulic Design DSRS 4.4, Rev 0: Thermal and DSRS Hydraulic Design 4.4, Rev 0: Thermal and DSRS Hydraulic Design
Tier 2 1.9-76 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.1 4.2 4.3 4.6.2 4.6.2 4.6.2 Comments Section NuScale does not interpret26 as GDC of means two safety-related requiring reactivity control. One of the independent reactivity control systems used to meet the NuScale26 in the of GDC requirements the chemical and volume control design is system, which is not safety-related. design-specific Principal Design Criterion in as reflected GDC 27, in lieu of (PDC) Section 3.1. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms 4.6.2 Conforms None. 4.6.2 Standard (DSRS) (Continued) AC Title/Description Conformance Environmental and Dynamic Effects - Dynamic Effects Environmental and GDC 4 Modes Failure and Effects 23 - GDC ConformsSingle Malfunction - GDC 25 None.OperationalControl and Reliability - GDC 26 Conforms None.Combined- GDC 27 Capability 28- GDC Reactivity Limits Departure NuScale The design bases conform to a Anticipated Protection Against 4.6.2 29 Occurrences - GDC Operational Conforms None. 4.6.2 4.6.0.2 AC II.2 II.3II.4 Nondestructive ExaminationII.5 Austenitic Stainless SteelsII.1 Conforms Other Materials None. Conforms None. Conforms None. 4.5.2 4.5.2 4.5.2 II.5 II.3 II.4 II.6 II.7 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 4.6, Rev 2: Functional Drive Control Rod of Design System SRP 4.5.2, Rev 3: Reactor 4.5.2, Rev 3: Reactor SRP and Core Support Internal Materials Structure SRP 4.5.2, Rev 3:Reactor SRP and Core Support Internal Materials Structure 4.5.2, Rev 3: Reactor SRP and Core Support Internal Materials Structure SRP 4.6, Rev 2: Functional Drive Control Rod of Design System SRP 4.6, Rev 2: Functional Drive Control Rod of Design System SRP 4.6, Rev 2: Functional Drive Control Rod of Design System SRP 4.6, Rev 2: Functional Drive Control Rod of Design System SRP 4.6, Rev 2: Functional Drive Control Rod of Design System SRP 4.6, Rev 2: Functional Drive Control Rod of Design System
Tier 2 1.9-77 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.2.2 5.2.2 Not Applicable Not Applicable Not Applicable is does not assume not assume is does Comments Section referencedcriterion in this acceptance are applicablenot or only partially applicable. This guidance is applicable only to BWR plants. designsto PWR only BTP is applicable This Control Offset Axial Constant use the that notdoesoperating scheme. NuScale use Control operating Offset Axial Constant the scheme. This guidance is applicable only to BWR plants. The overpressure analys a secondary safety-grade signalthe from NuScale does the reactor trip. RPS initiates safety-grade reactor have a secondary not system. trip ormssubtier guidance documents Certain Status Review Plan (SRP) and Design Specific Review Conforms in TableRG 1.26 See 1.9-2.Conforms in TableRG 1.26 See 1.9-2. 5.2.1 5.2.1 Conforms None. 5.2.2 Not Applicable Partially Conforms Partially in TableRG 1.147 See 1.9-2.Conforms Partially in TableRG 1.192 See 1.9-2.Partially Conforms 5.2.1 5.2.1 Standard (DSRS) (Continued) AC Title/Description Conformance BWR Alternate Rod Injection System Injection Rod Alternate BWR Applicable Not and 1 1.26 to meet GDC RG of Use 10 CFR 50.55a and 1 1.84 to meet GDC RG of Use 10 CFR 50.55a Use of RG 1.147 to meet GDC and 1 10 CFR 50.55a Use of RG 1.192 to meet GDC and 1 10 CFR 50.55a BWRs Design Requirements for Operating at Power PWRs Design Requirements for Operating at Power PWRs Design Requirements for at Low TemperatureOperating (Startup, Shutdown) AC II.8 AllII II.1 -II.2 II.3 II.1II.2 Applicable Not II.3 Material SpecificationsII.4 ConformsII.5 None.II.6 Testing and Inspections Technical Specifications ConformsConf Partially None. 5.2.2 5.2.2 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 4.6, Rev 2: Functional Drive Control Rod of Design System 3: Westinghouse Rev BTP 4-1, AxialConstant Offset Control (CAOC) 5.2.1.1, Rev 3: SRP with the Codes Compliance Rule, Standards and 10 CFR 50.55a 5.2.1.2, Rev 3: Applicable SRP Code Cases 5.2.1.2, Rev 3: Applicable SRP Code Cases 5.2.1.2, Rev 3: Applicable SRP Code Cases Rev 3: Overpressure SRP 5.2.2, Protection Rev 3: Overpressure SRP 5.2.2, Protection Rev 3: Overpressure SRP 5.2.2, Protection Rev 3: Overpressure SRP 5.2.2, Protection Rev 3: Overpressure SRP 5.2.2, Protection Rev 3: Overpressure SRP 5.2.2, Protection
Tier 2 1.9-78 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.2.3 5.2.3 5.2.3 5.2.3 5.2.3 Comments Section This acceptance criterion is applicable This subtier guidance except for references to that is applicableBWRs. only to acceptance criterion is applicable This subtier guidance except for references to that is applicableBWRs. only to acceptance criterion is applicable This subtier guidance except for references to or to large BWRs to is applicable only that LWRs that use nonmetallic thermal insulation on reactor coolant pressure boundary components. acceptance criterion is applicable This subtier guidance except for references to that is applicableBWRs. only to acceptance criterion is applicable This subtier guidance except for references to that is applicableBWRs, only and to to 1.37, which endorses use of NQA- subtier RG The NuScale design is based on 1-1994. and the NQA-1a-2009 NQA-1-2008 addenda, rather than NQA-1-1994. Status Review Plan (SRP) and Design Specific Review Conforms None. 5.2.3 Conforms None. 5.2.3 Partially Conforms Partially Conforms Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance TMI Action Plan Requirements Action TMI Conforms None.- 10Fracture Toughness CFR 50, G Appendix Steel Tubular ProductsNDE of Ferritic Conforms None. 5.2.2 GDC 4 Compatibility of Components - Measures to Avoid Sensitization in Austenitic Stainless Steel GDC 4 Compatibility of Components - Controls to Avoid Stress Corrosion Cracking in Austenitic Stainless Steel 5.2.3 Reactor Coolant Materials Austenitic Stainless Steel AC II.7 II.1II.2 Material SpecificationsII.3 Compatibility of Materials with the II.3.A Conforms Partially Fabrication and Processing of Ferritic II.3.BII.3.C WeldingFerritic Steel Control of II.4 Conforms None. Fabrication and Processing of II.4.A II.4.B 5.2.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 5.2.2, Rev 3: Overpressure SRP 5.2.2, Protection 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials
Tier 2 1.9-79 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.2.3 Not Applicable Not Applicable Comments Section This acceptance criterion is applicable only This to LWRs that use nonmetallic thermal insulation on reactor coolant pressure boundary components. NuScale does not use nonmetallic thermal insulation on reactor coolant pressureboundary components. acceptance criterion is applicable This subtier guidance except for references to that is applicableBWRs. only to acceptanceThis criterion governs plant- is specific programmatic information that the responsibilityCOL of the applicant. Status Review Plan (SRP) and Design Specific Review Conforms None. 5.2.3 Conforms None. 5.2.4 Not Applicable Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance NDE of Austenitic StainlessSteel Tubular Products Steel Materials with Thermal Insulation Stainless Steels Inspection AC II.4.C Stainless of Austenitic Compatibility II.4.DII.4.E Austenitic Control Welding of of II.4.G Operational Programs Not Applicable II.1II.2 to Subject Boundary System II.3 AccessibilityII.4 Examination Categories and Methods ConformsII.5 Inspection Intervals None. Conforms Evaluation of ExaminationResults None. Conforms Conforms None. None. 5.2.4 5.2.4 5.2.4 5.2.4 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 5.2.3, Rev 3: Reactor 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials 5.2.3, Rev 3: Reactor SRP Coolant Pressure Boundary Materials DSRS 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing
Tier 2 1.9-80 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.2.4 5.2.4 Not Applicable Not Applicable Comments Section Although a boric acidhas control program been established,not a brief description of provided in the DCA. is the program acceptanceThis criterion governs plant- is specific programmatic information that the responsibilityCOL of the applicant. criterionis of this acceptanceA portion applicableonly to COL applicants. acceptanceThis criterion governs plant- is specific programmatic information that the responsibilityCOL of the applicant. Status Review Plan (SRP) and Design Specific Review Conforms None. 5.2.4 Standard (DSRS) (Continued) ograms Partially Conforms AC Title/Description Conformance Augmented ISI to Protect Against Protect ISI to Augmented Piping Failures Postulated ITAAC ProgramInformed ISI Risk Applicable Not Partially Conforms AC II.11 Pr Inspection Other II.6II.7 System Pressure TestsII.8 Code Exemptions ConformsII.9 Relief Requests None.II.10 Cases Code Conforms None. Conforms None. Conforms None. 5.2.4 5.2.4 5.2.4 5.2.4 II.12 Operational ProgramsII.13 II.14 Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing DSRS 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing DSRS 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing DSRS 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing 5.2.4, Rev 0: Reactor DSRS Coolant Pressure Boundary Inservice Inspection and Testing
Tier 2 1.9-81 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None.Conforms None. 5.3.1 5.3.1 5.3.1 5.3.2 Standard (DSRS) (Continued) AC Title/Description Conformance Criteria to Meet GDC 2Criteria to Meet GDC 14Criteria to Meet GDC 30Criteria to Meet Conforms None. Conforms None. Conforms None. 5.2.5 5.2.5 5.2.5 Manufacture and Fabrication of and Manufacture Components Examination and Steels for Ferritic Used Processes Austenitic Stainless Steels Regulations, Codes,Basis and Documents AC II.1 II.2 II.3 II.1II.2 MaterialsII.3for Special Processes Used II.4 Special Methods for Nondestructive II.5 Special Controls and Special II.6II.7 Fracture Toughness ConformsII.1.A Material Surveillance None.Fasteners Reactor Vessel Pressure-Temperature - Applicable II.1.B Conforms Pressure-Temperature Requirements Conforms None. Conforms None. Conforms None. None. 5.3.1 5.3.1 5.3.1 5.3.1 5.3.2 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 5.2.5, Rev 0: Reactor DSRS Coolant Pressure Boundary Leakage Detection DSRS 5.2.5, Rev 0: Reactor DSRS Coolant Pressure Boundary Leakage Detection 5.2.5, Rev 0: Reactor DSRS Coolant Pressure Boundary Leakage Detection 5.3.1, Rev 0: Reactor DSRS Vessel Materials 5.3.1, Rev 0: Reactor DSRS Vessel Materials 5.3.1, Rev 0: Reactor DSRS Vessel Materials 5.3.1, Rev 0: Reactor DSRS Vessel Materials 5.3.1, Rev 0: Reactor DSRS Vessel Materials 5.3.1, Rev 0: Reactor DSRS Vessel Materials 5.3.1, Rev 0: Reactor DSRS Vessel Materials 5.3.2, Rev 0: Pressure- DSRS Temperature Limits, Upper- Shelf Energy, and Pressurized Thermal Shock 5.3.2, Rev 0: Pressure- DSRS Temperature Limits, Upper- Shelf Energy, and Pressurized Thermal Shock
Tier 2 1.9-82 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.3.2 5.3.3 Not Applicable e reactor vessel is Comments Section e surveillance of th This acceptanceThis criterion governs plant- is specific programmatic information that the responsibilityCOL of the applicant. described in the DCD. However, the COL described in the DCD. However, applicant develops and implements the reactor vessel surveillanceprogram. Status Review Plan (SRP) and Design Specific Review Conforms None. 5.3.2 Conforms None.Conforms None. 5.3.2 5.3.2 Standard (DSRS) (Continued) AC Title/Description Conformance Regulations, Codes,Basis and Documents Applicable Regulations, Codes, and Documents Basis Requirements AC II.2.AApplicable Upper-Shelf Energy - II.2.B Upper-Shelf Energy Requirements Conforms None. 5.3.1 II.3.A Pressurized Thermal Shock - II.3.B Pressurized Thermal Shock II.1II.2II.3 DesignII.4 Materials of ConstructionII.5 Fabrication MethodsII.6 Inspection RequirementsII.7 Shipment and Installation Conforms Operating Conditions None. Conforms Conforms Inservice Surveillance Conforms None. Conforms None. None. None. Conforms None. Conforms Inservic 5.3.3 5.3.3 5.3.3 5.3.3 5.3.3 5.3.3 II.8 Operational Programs Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 5.3.2, Rev 0: Pressure- DSRS Temperature Limits, Upper- Shelf Energy, and Pressurized Thermal Shock DSRS 5.3.2, Rev 0: Pressure- DSRS Temperature Limits, Upper- Shelf Energy, and Pressurized Thermal Shock DSRS 5.3.2, Rev 0: Pressure- DSRS Temperature Limits, Upper- Shelf Energy, and Pressurized Thermal Shock 5.3.2, Rev 0: Pressure- DSRS Temperature Limits, Upper- Shelf Energy, and Pressurized Thermal Shock 5.3.3, Rev 0: Reactor DSRS Vessel Integrity 5.3.3, Rev 0: Reactor DSRS Vessel Integrity 5.3.3, Rev 0: Reactor DSRS Vessel Integrity 5.3.3, Rev 0: Reactor DSRS Vessel Integrity 5.3.3, Rev 0: Reactor DSRS Vessel Integrity 5.3.3, Rev 0: Reactor DSRS Vessel Integrity 5.3.3, Rev 0: Reactor DSRS Vessel Integrity DSRS 5.3.3, Rev 0: Reactor DSRS Vessel Integrity
Tier 2 1.9-83 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.4.1 5.4.1 5.4.1 Not Applicable Not Applicable Comments Section btier guidance documents This requirement applies to plant-specificThis the responsibility of is verification and applicant. COL criteriasection and its acceptance SRP This designs PWR apply only to through II.7) (II.1 that use reactor coolant pumps. The not require or NuScale reactor design does include reactor coolant pumps. Rather, the NuScale design uses passive natural circulation of the primary coolant, eliminating the need for reactor coolant pumps. criterionis of this acceptanceA portion applicableto COL applicants referencing a certified design. referencedcriterion in this acceptance are only partially applicable. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 5.4.1 5.4.1 5.2 Conforms None.Conforms None.Conforms None. 5.4.1 5.4.1 5.4 Partially Conforms Certain su Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance 10 CFR compliance 52.47(b)(1) Not Applicable Inspection of Materials Materials Austenitic Stainless Steel in Technical Specifications Primary (Reactor) and Secondary Cleanliness Control Coolant and Side of the Steam Secondary Generator to Degradation Program Elements AC II.9 All Various Not Applicable II.1II.2II.3 Selection, Processing, Testing, and II.4 Steam Generator Design Fabrication and Processing of Ferritic Fabrication and Processing of Conforms None.II.3 Steam Generator Program Elements 5.4.1 II.5II.6 Compatibility of Materials with the II.1 Provisions for Accessing the II.2 Susceptibility Steam Generator Tube Steam Generator Monitoring II.4Criteria Steam Generator Tube Repair Conforms None. 5.4.1 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 5.3.3, Rev 0: Reactor DSRS Vessel Integrity 5.4.1.1, Rev 3: Pump SRP Flywheel Integrity (PWR) DSRS 5.4.2.1, Rev 0: Steam Generator Materials DSRS 5.4.2.1, Rev 0: Steam Generator Materials DSRS 5.4.2.1, Rev 0: Steam Generator Materials DSRS 5.4.2.1, Rev 0: Steam Generator Materials DSRS 5.4.2.2, Rev 0: Steam Generator Program DSRS 5.4.2.1, Rev 0: Steam Generator Materials DSRS 5.4.2.1, Rev 0: Steam Generator Materials DSRS 5.4.2.2, Rev 0: Steam Generator Program DSRS 5.4.2.2, Rev 0: Steam Generator Program DSRS 5.4.2.2, Rev 0: Steam Generator Program
Tier 2 1.9-84 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.4.3 5.4.1 5.4.1 Not Applicable Not Applicable Comments Section This SRP section and its acceptance criteriasection and its acceptance SRP This BWRs. apply only to through II.10) (II.1 system andinterface with directly does not the RCPB. provisionsthe powerfrom of GDC 34. As described in Sectionthe design 3.1.4, NuScale-specific principal complies with a design criterion in lieuGDC. of this This acceptanceThis criterion governs plant- activitiesthe thatspecific programmatic are applicantresponsibilitythe COL of referencing a certified design. applicableonly to COL applicants. mscriterionis of this acceptance A portion Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 5.4.1 5.4.1 Partially Conforms 5.4.1 Standard (DSRS) (Continued) AC Title/Description Conformance Methods Inspection Certified Design Technicalin Specifications AC II.5II.6II.7 Repair Steam Generator Tube Steam Generator Tube Preservice and Testing Periodic Tube Inspection All Various Not Applicable II.1 thru II.3 VariousII.4II.5 GDC 5II.6 GDC 14II.7 GDC 19 Conforms None. GDC 34 Conforms Applicable Not None. the secondary to is connected The DHRS Conforms None. Departuredesign supports an exemption NuScale The 5.4.3 5.4.3 5.4.3 II.8II.9 Operational Programs ITAAC Partially Conforms Confor Partially Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 5.4.2.2, Rev 0: Steam Generator Program Core 5.4.6, Rev 4: Reactor SRP Isolation Cooling System (BWR) DSRS 5.4.2.2, Rev 0: Steam Generator Program DSRS 5.4.2.2, Rev 0: Steam Generator Program DSRS 5.4.7, Rev 0: Decay Heat 5.4.7, Rev 0: Decay DSRS Removal (DHR) System Responsibilities Heat 5.4.7, Rev 0: Decay DSRS Removal (DHR) System Responsibilities Heat 5.4.7, Rev 0: Decay DSRS Removal (DHR) System Responsibilities Heat 5.4.7, Rev 0: Decay DSRS Removal (DHR) System Responsibilities Heat 5.4.7, Rev 0: Decay DSRS Removal (DHR) System Responsibilities DSRS 5.4.2.2, Rev 0: Steam Generator Program DSRS 5.4.2.2, Rev 0: Steam Generator Program
Tier 2 1.9-85 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.4.3 Not Applicable Not Applicable Comments Section closed-loop DHRS outside the containment is directly connected to the containment is directly connected closed-loopRPV system within the SG providing dual passive barriers between the NPM. the the reactor pool outside and RCS system outsideBreaches of this piping considered credible containment is not because the system is a welded design with a system design pressure equivalent to the RPV, designed to Class 2 requirements in Section BPV Code, accordance with ASME criteria of NRC meets the applicable III, and Branch Technical Position 3-4, Revision 2. As isolationleakage detection and a result, capabilities of this piping system from considered important containment are not to safety. This SRP section and its acceptance criteriasection and its acceptance SRP This BWRs. apply only to through II.4) (II.1 criteriasection and its acceptance SRP This use a that to PWRs only and II.2) apply (II.1 pressurizer reliefpressurizer relief tank. A tankisnot used in the NuScaledesign. Fluid relieved through the reactor coolant system protection system is routed overpressure directly to the containment vessel. Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance AC II.8 GDC 54Conforms Partially This AllAll Various Various Not Applicable Not Applicable II.9with other systems DHRS Interface Conforms None. 5.4.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 5.4.7, Rev 0: Decay Heat 5.4.7, Rev 0: Decay DSRS Removal (DHR) System Responsibilities 5.4.8, Rev 3: Reactor SRP Water Cleanup System (BWR) 5.4.11, Rev 4: Pressurizer SRP Tank Relief DSRS 5.4.7, Rev 0: Decay Heat 5.4.7, Rev 0: Decay DSRS Removal (DHR) System Responsibilities
Tier 2 1.9-86 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.4.1 5.4.1 10.3.5 10.3.5 10.3.5 Not Applicable Not Applicable Not Applicable Comments Section bstantively similar system configuration, non-condensable spacegases in the pressurizer accumulating will not interfere with core cooling during or after designaccidents. basis The NuScale design supportsan exemption from the requirementsof 10 CFR 50.46a related to reactor coolant system high point venting, as well the su requirements of 10 CFR 50.34(f)(2)(vi). This SRP section and its acceptance criteriasection and its acceptance SRP This to are applicable only through II.12) (II.1 BWRs. criterionof this acceptanceA portion governs information that is site-specific and is theresponsibilityof the COL applicant referencing the certified design. criterionof this acceptanceA portion governs information that is site-specific and is theresponsibilityof the COL applicant referencing the certified design. acceptance criterion is the This applicantresponsibilitythe COL of referencing the certified design. Status Review Plan (SRP) and Design Specific Review Conforms None. 5.2.2 Conforms None. 5.4.1 Not Applicable Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance System Design, Installation, and Installation, Design, System Capabilities to Prevent Exceeding andTechnical NRC Specifications Requirements Regulatory Meeting Industry Guidelines Meeting Industry Parameters Reporting AC All VariousAllB.1 Various DepartureB.2Program Secondary Water Chemistry of the integral reactor coolant Because B.3 Sampling Schedule for Critical B.4 Records Not Applicable B.1 Program Change Evaluation and Partially Conforms Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 5.4.12, Rev 1: Reactor SRP Point High System Coolant Vents 5.4.13, (March 2007): SRP Isolation Condenser System (BWR) of 3: Monitoring Rev BTP 5-1, Secondary Side Water Steam Chemistry in PWR Generators of 3: Monitoring Rev BTP 5-1, Secondary Side Water Steam Chemistry in PWR Generators of 3: Monitoring Rev BTP 5-1, Secondary Side Water Steam Chemistry in PWR Generators of 3: Monitoring Rev BTP 5-1, Secondary Side Water Steam Chemistry in PWR Generators 3: Rev BTP 5-2, Overpressurization Protection of Pressurized- Water Reactors While Operating at Low Temperatures
Tier 2 1.9-87 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.2.2 Ch 16 Comments Section ASME Section XI Appendix G Criteria. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 5.2.2 5.2.2 Conforms None. 5.2.2 Partially Conforms Conforms to sting Conforms None. 5.2.2 Standard (DSRS) (Continued) AC Title/Description Conformance Protection Operability Active Component Failure Design Operating-Basis Earthquake AC B.2 Low-Temperature Overpressure B.3 Single Withstand to Designed System B.4Instrumentation and Controls System B.5 System Operability Te B.6 Applicable GuidanceB.7 to Design System Withstand Conforms None. 5.2.2 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: BTP 5-2, Rev 3: Rev BTP 5-2, Overpressurization Protection of Pressurized- Water Reactors While Operating at Low Temperatures 3: Rev BTP 5-2, Overpressurization Protection of Pressurized- Water Reactors While Operating at Low Temperatures 3: Rev BTP 5-2, Overpressurization Protection of Pressurized- Water Reactors While Operating at Low Temperatures 3: Rev BTP 5-2, Overpressurization Protection of Pressurized- Water Reactors While Operating at Low Temperatures 3: Rev BTP 5-2, Overpressurization Protection of Pressurized- Water Reactors While Operating at Low Temperatures 3: Rev BTP 5-2, Overpressurization Protection of Pressurized- Water Reactors While Operating at Low Temperatures
Tier 2 1.9-88 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.3 5.3 5.3 5.2.2 5.2.2 fromthe primary ptance criterion apply pplicable except for Comments Section ce criterion is applicable The intent of this guidance - that the low The intent temperature overpressure protection (LTOP) system should not depend on the of offsite power to perform itsavailability function - appliesdesign. to the NuScale isof this acceptance criterion The intent but the criterion refers to large applicable LWR designs that provide pressure relief normally not system a low-pressure from primarythe system. Into the connected NuScale design, the LTOP system is not to a low-pressureconnected system. - to this guidance intent of the However, system is not LTOP the ensure that inadvertently isolated system - is applicablethe DCA. to Portions of this acce plants for which fracture older only to toughness testing on vessel material did includenot all tests necessary to determine of this guidance applies to The rest RTNDT. the NuScale design. This guidance is a which reference to subtier NUREG-0744, applies only to operating reactors that do minimum fracturemeet the toughness not acceptancecriteria defined in this BTP 5-3. exceptinindicated as the commentsbelow Acceptance Criteria for 1.1 and 1.2. Status Review Plan (SRP) and Design Specific Review Conforms None. 5.2.2 Partially Conforms Partially Conforms Partially Conforms This acceptan Partially Conforms forVessel Standard (DSRS) (Continued) NDT AC Title/Description Conformance Determination of RT Materials System Actuation Protection Requirements Shelf Energies AC B.8 Backup Electrical Power SourceB.9 PartiallyConforms Analyses Considering Inadvertent B.10 Interlocks to Ensure Overpressure 11.1 Test Preservice Toughness Fracture 1.2UpperV-Notch Estimation of Charpy Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: BTP 5-2, Rev 3: Rev BTP 5-2, Overpressurization Protection of Pressurized- Water Reactors While Operating at Low Temperatures 3: Rev BTP 5-2, Overpressurization Protection of Pressurized- Water Reactors While Operating at Low Temperatures 3: Rev BTP 5-2, Overpressurization Protection of Pressurized- Water Reactors While Operating at Low Temperatures 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements
Tier 2 1.9-89 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.3 Not Applicable Not Applicable Comments Section criterion applies except as This acceptanceThis criterion governs plant- specific reporting criteria that are the responsibility of the COL applicant that referencesNuScale certified the design. acceptanceThis criterion governs plant- specific activities that are the responsibility the that references the COL applicant of NuScale certified design. indicated in the comments below for 3.5. 3.4 and Acceptance Criteria Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 5.3 5.3 5.3 Conforms None. 5.3 Partially Conforms This acceptance Standard (DSRS) (Continued) AC Title/Description Conformance SAR Criteria Conforms None. 5.3 Toughness Limitations Limitations Toughness Requirements AC 1.322.1 Reporting Requirements2.2Limitations Operating for Fracture 2.3 Pressure-Temperature Operating 3 Recommended Bases for Operating Conforms Reporting Requirements None.3.1 Inservice Surveillance of Fracture 3.2 3.3Requirements Surveillance Program Conforms3.4 Conforms None. Surveillance Test Procedures None.3.5 Reporting Criteria Conforms4.1 Technical Specification Changes 5.3 None.B.1 Not Applicable Not Applicable Pressurized Thermal Shock B.2 Functional Requirements Pressure Relief Requirements 5.3 5.3 Conforms Conforms None. None. 5.3 5.4.3 5.4.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: BTP 5-3, Rev 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 2: Fracture Rev BTP 5-3, Toughness Requirements 5-4, Rev 0: Design BTP DSRS Requirements of the Residual Heat Removal System 5-4, Rev 0: Design BTP DSRS Requirements of the Residual Heat Removal System
Tier 2 1.9-90 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 2.4 19.2 5.4.3 6.1.1 3.8.2 3.9.3 6.2.1 6.2.1 Not Applicable on surfaces inside pplicable except the Comments Section The procedures governed byThe procedures this governed acceptancecriterion are site-specific and are the responsibilityCOL of the applicant. This guidance is a NuScale design does not provide a method post-accidentfor addressed pH control as in BTP 6-1. only to the section is applicable SRP This use of protective coatings the containment. The NuScale Power protectiveModule design does not use coatings inside the containment vessel. See SRP 2.4.6, 2.4.10, 2.4.12, 3.9.3, 19.0, and 3.9.3, 2.4.12, 2.4.6, 2.4.10, SRP See 3.8.2. DSRS limiting eventlimiting is less than thedesign pressure. ms has been withdrawn by the NRC. RG 1.37 6.1.1 Status Review Plan (SRP) and Design Specific Review Conform peakthe The containment pressure for Conforms SRP or DSRS. SRP See theapplicable Standard (DSRS) (Continued) AC Title/Description Conformance Composition and Compatibility of and Compatibility Composition ESF Systems Fluids Applicable acceptance criteria are 2.4.12, 2.4.6, 2.4.10, addressed in SRP 3.8.2. and DSRS 3.9.3, 19.0 Design Pressure Design AC B.3B.4 Test RequirementsB.5 Operational ProceduresII.1 ImplementationII.1.A Partially Conforms ConformsII.1.B Materials and Fabrication None. Austenitic Stainless SteelsII.2 Ferritic Steel Welding Conforms Conforms Conforms None. None. None. Conforms None. 5.4.3 5.4.3 6.1.1 6.1.1 6.1.1 II.3 Component Cleaning and Systems PartiallyConfor No specific requirements listed. II.4All Thermal Insulation Various Conforms None. Not Applicable 6.1.1 II.1Containment Design Margin for Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS BTP 5-4, Rev 0: Design BTP DSRS Requirements of the Residual Heat Removal System 5-4, Rev 0: Design BTP DSRS Requirements of the Residual Heat Removal System 5-4, Rev 0: Design BTP DSRS Requirements of the Residual Heat Removal System SRP 6.1.1, Rev 2: Engineered SRP Features Materials Safety 6.1.1, Rev 2: Engineered SRP Features Materials Safety 6.1.1, Rev 2: Engineered SRP Features Materials Safety 6.1.1, Rev 2: Engineered SRP Features Materials Safety SRP 6.1.1, Rev 2: Engineered SRP Features Materials Safety 6.2.1, Rev 0: DSRS Containment Functional Design SRP 6.1.1, Rev 2: Engineered SRP Features Materials Safety SRP 6.1.2, Rev 3: Protective 6.1.2, Rev 3: Protective SRP - (Paints) Systems Coating Materials Organic DSRS 6.2.1.1.A, Rev 0: Rev 0: 6.2.1.1.A, DSRS Containment
Tier 2 1.9-91 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.2.2 6.2.2 Not Applicable Not Applicable Not Applicable pplicable only to LWR subcompartments. Suppression Type Comments Section The NuScale design does not use an iceThe NuScale design does not usean condenser containment. criteriasection and its acceptance SRP This apply only to applicants for BWR designs that involve Pressure Containments. criteriasection and its acceptance SRP This (II.1 through II.4) are a containment designs that involve a structure thathouses design vessel The NuScale containment does not have subcompartments housing high-energy pipingin this as defined internalguidance (or compartments as referred to in GDC 50). Status Review Plan (SRP) and Design Specific Review Conforms None. 6.2.1 Conforms None. 6.2.1 Conforms None. 6.2.1 Conforms None.Conforms None.Conforms None.Conforms None.Conforms None. 6.2.1 6.2.1 6.2.1 6.2.1 6.2.1 Standard (DSRS) (Continued) AC Title/Description Conformance ContainmentRemoval Heat - LOCA Design Margin and Capability Assumptions FollowingDesign Basis Postulated Accident Capability and Design Margin - Containment Response Analysis Assumptions External Pressure Conditions Instrumentation Structures and System Components Generated Hydrogen Modules AC II.2II.3 Reducing Containment Pressure AllAll VariousAll Various Various Not Applicable Not Applicable Not Applicable II.4Removal Containment Heat II.5II.6II.7 Protection of Containmentfrom II.8 Containment Monitoring II.9 Design of Containment Internal Involving of Accident Evaluation Evaluation of an Accident on other Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 6.2.1.1.A, Rev 0: Rev 0: 6.2.1.1.A, DSRS Containment Rev 3: Ice 6.2.1.1.B, Draft SRP Condenser Containments SRP 6.2.1.1.C, Rev 7: Pressure Suppression BWR Type Containments 6.2.1.2, Rev 3: SRP Subcompartment Analysis DSRS 6.2.1.1.A, Rev 0: Rev 0: 6.2.1.1.A, DSRS Containment DSRS 6.2.1.1.A, Rev 0: Rev 0: 6.2.1.1.A, DSRS Containment DSRS 6.2.1.1.A, Rev 0: Rev 0: 6.2.1.1.A, DSRS Containment Rev 0: 6.2.1.1.A, DSRS Containment Rev 0: 6.2.1.1.A, DSRS Containment Rev 0: 6.2.1.1.A, DSRS Containment Rev 0: 6.2.1.1.A, DSRS Containment
Tier 2 1.9-92 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.2.1 9.2.5 Not Applicable Not Applicable Comments Section This acceptance criterion is applicable only This to COL applicants. criteriasection and its acceptance SRP This a for which PWRs only to are applicable uncovery. in core results LOCA postulated design, a LOCA For the NuScale reactor result in core uncovery. does not not included. Theincluded. NuScale design supports not an exemption from selected portions of 10 K. 50, Appendix CFR Status Review Plan (SRP) and Design Specific Review Conforms None. 6.2.1 Conforms None. 6.2.2 Departureis from metal-water reactions energy The Not Applicable Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Compliance with GDC 50 and 10 CFR 50, Appendix< K paragraph I.A of Heat during the LOCA - Sources Inspections, Tests, Analyses, and for Criteria Acceptance (ITAAC) Design Certification Applications Inspections, Tests, Analyses, and for Criteria Acceptance (ITAAC) Combined(COL) Applications License Containment Pressure Model for Analysis; ECCS Performance Containment Response Analyses Conservatism of Structures, Sharing GDC 5, ComponentsSystems, and AC II.1 II.2 II.3 All II.1II.2 Sources of EnergyII.3Rate Release and Energy Mass Single-Failure Analyses Conforms Conforms None. None.II.1 Conforms None. 6.2.1 6.2.1 6.2.1 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 6.2.1.3, Rev 0: Mass and 6.2.1.3, DSRS Release Analysis for Energy Loss-of-Coolant Postulated Accidents (LOCAs) Rev 0: Mass and 6.2.1.3, DSRS Release Analysis for Energy Loss-of-Coolant Postulated Accidents (LOCAs) Rev 0: Mass and 6.2.1.3, DSRS Release Analysis for Energy Loss-of-Coolant Postulated Accidents (LOCAs) 6.2.1.5, Rev 3: Minimum SRP Containment Pressure Core for Emergency Analysis Cooling System Performance Capability Studies DSRS 6.2.1.4, Rev 0: Mass and 6.2.1.4, DSRS Release Analysis for Energy Postulated Secondary System Pipe Ruptures Rev 0: Mass and 6.2.1.4, DSRS Release Analysis for Energy Postulated Secondary System Pipe Ruptures Rev 0: Mass and 6.2.1.4, DSRS Release Analysis for Energy Postulated Secondary System Pipe Ruptures 6.2.2, Rev 0: DSRS ContainmentRemoval Heat Systems
Tier 2 1.9-93 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.2.4 6.2.2 3.1.4 6.2.2 Not Applicable Comments Section This SRP section and its acceptance criteriasection and its acceptance SRP This designs LWR apply only to through II.4) (II.1 that incorporate primary and secondary containmentcontainment. The NuScale vessel design does not include a secondary containment. containment. The NuScale design supports an exemptionThe NuScale provisionsthe powerfrom of GDC 38. As described in Sectionthe design 3.1.4, NuScale-specificcomplies with a principal design criterion in lieuGDC. of this designand supports the GDC 40 an exemption. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 6.2.2 6.2.2 6.2.2 Conforms None.Conforms None.Conforms None. 6.2.4 6.2.4 6.2.4 Departure NuScale The design does not conform to Standard (DSRS) (Continued) AC Title/Description Conformance as it relates to requirements for long- to requirements for as it relates term cooling GDC 38, Containment Heat Removal Departure Containment of Inspection GDC 39, Heat Removal System GDC 40, Testing of Containment Heat Removal System 10 CFR cooling, long-term 50.46(b)(5), including adequate water level (head) margin RRVs), in the presence of LOCA-generated and latent debris Lines in Engineered Safety Feature (or Systems Related) Systems Needed for Safe in Lines Shutdown Requirements AC II.4 II.6 with 10 Compliance CFR 50.46(b)(5) II.2 All Various Not Applicable II.1II.2 Instrument Line IsolationII.3Detection in Leak Isolation of and II.4Detection in Leak Isolation of and Conformsinstrumentation No process lines penetrate Isolation Valve Containment II.3 II.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 6.2.2, Rev 0: DSRS ContainmentRemoval Heat Systems 6.2.2, Rev 0: DSRS ContainmentRemoval Heat Systems DSRS 6.2.2, Rev 0: DSRS ContainmentRemoval Heat Systems 6.2.3, Rev 3: Secondary SRP Containment Functional Design DSRS 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System DSRS 6.2.2, Rev 0: DSRS ContainmentRemoval Heat Systems 6.2.2, Rev 0: DSRS ContainmentRemoval Heat Systems
Tier 2 1.9-94 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section ablevalves are not used as CIVs. Relief Not Applicable Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 6.2.4 6.2.4 6.2.4 Conforms None.Conforms None. 6.2.4 6.2.4 Conforms None. 6.2.4 Standard (DSRS) (Continued) ion Valves Outside Automatic Isolation Isolation Automatic AC Title/Description Conformance Containment Containment Isolation Valves Requirements for Engineered Safety Feature (or Related) Systems Valves Isolation Automatic of NonEssential Systems AC II.10II.11to Loss of Power II.12 Isolation Reliabilityof Diversity for Initiation Parameter Conforms None. 6.2.4 II.5II.6 Isolation Valve Containment II.7 Barriers Use of inSealed-Closed Place Use of Relief Valves as IsolationValves Not Applic II.8II.9of Essential Classification or of Isolat Location Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System DSRS 6.2.4, Rev 0: DSRS Containment Isolation System DSRS 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System DSRS 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System
Tier 2 1.9-95 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.2.4 )(2)(xiv). om 50.34(f Comments Section potential for an open path from but is isolated containment to the environs containmentuponvessel pressure a high and a low-low pressurizersignal, level signal. Any in-containment event resulting in core damage or degradation also results in containment isolation on low pressurizer levelhigh containment and leading to core pressure. Any event damage or degradation, results in containment isolationlow on low features provide an pressurizer level. These alternative, reliable means to prevent releaseradiologicalCES to the from the consistentenvirons,of this with the intent AcceptanceThe NuScaledesign Criterion. supports an exemption fr Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None.Conforms None. 6.2.4 6.2.4 6.2.4 6.2.4 Departure The containment evacuation system has the Standard (DSRS) (Continued) AC Title/Description Conformance Containment Isolation on Open Paths to the Environs Containment Containment Isolation Components Operator Actions Valves AC II.13for Initiation of Radiation Monitors II.14II.15 Isolation Valve Closure TimesII.16 Use of Closed System Inside ConformsII.17Design Criteria for Specific None.II.18 Provisions to Allow Control Room II.19Testing OperabilityLeakage Rate and Conforms Containment Isolation of Reopening None. 6.2.4 6.2.4 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 6.2.4, Rev 0: DSRS Containment Isolation System DSRS 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.4, Rev 0: DSRS Containment Isolation System
Tier 2 1.9-96 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.2.5 6.2.5 Not Applicable Comments Section not include a containment atmosphericnot integrity is cleanup system. Containment to control systems without assured oxygen concentrationshydrogen and within containment. acceptance aboveGDC for criterion II.4 See compliance. 41 from the power provisionsthe powerfrom of GDC 41. As described in Sectionthe design 3.1.4, NuScale-specific principal complies with a design criterion in lieuGDC. of this concentrations within containment are not required because combustionhas no impact on CNV integrity. Status Review Plan (SRP) and Design Specific Review Conforms None. 6.2.5 Conforms None. 6.2.4 Conforms None. 6.2.5 Not Applicable For GDC 42 and the NuScale design does 43, PartiallyConforms Systems to control hydrogen Standard (DSRS) (Continued) AC Title/Description Conformance Analysis Design Requirements of GDC 41Design Requirements of DepartureTest Requirements of Inspection and design supports an exemption NuScale The 43 GDC 42, and GDC GDC 41, Calculations Concentration Control and Distribution in Containment Containment Structural Integrity AC II.6 Containment Structural Integrity II.20II.21 Station BlackoutII.1 Source Term in Radiological Analysis of Hydrogen and Oxygen Conforms None. 6.2.4 II.5 II.2II.3 Equipment Survivability and II.4 Ensuring a Mixed Atmosphere Conforms None. 6.2.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 6.2.5, Rev 0: DSRS Combustible Gas Control in Containment DSRS 6.2.4, Rev 0: DSRS Containment Isolation System DSRS 6.2.4, Rev 0: DSRS Containment Isolation System 6.2.5, Rev 0: DSRS Combustible Gas Control in Containment DSRS 6.2.5, Rev 0: DSRS Combustible Gas Control in Containment DSRS 6.2.5, Rev 0: DSRS Combustible Gas Control in Containment 6.2.5, Rev 0: DSRS Combustible Gas Control in Containment 6.2.5, Rev 0: DSRS Combustible Gas Control in Containment
Tier 2 1.9-97 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.2.6 6.3.1 6.3.3 Comments Section from the containment leakage ratethe containment leakagefrom testing design pressure requirements of GDCat 52 requirements of 10A test Type and CFR 50 Appendix J. The guidance in this acceptance criterion related to actuation signals is applicable to of ECCS actuation. For the requirements GDC 27, the NuScale ECCS does not perform a poison addition safety function nor does it function. The NuScaleprovide a makeup design supportsanGDC exemption to 27. the design described in Section 3.1.4, As NuScale-specific principal complies with a design criterion in lieuGDC. of this Status Review Plan (SRP) and Design Specific Review Conforms None. 6.3.1 Conforms None.Conforms None. 6.3.1 6.3.2 Departure None. 6.3.2 Departure Standard (DSRS) (Continued) AC Title/Description Conformance ECCS Acceptance of Criteria 10 CFR 50.46 of Non-Safety-Related Design Portions of ECCS SystemsECCS Interfaces and Shared Conforms on Reports and Times ECCS Outage None.Unavailability 6.3.1 Testing Capability and Actuation Provisions AC II.7 II.10 Programmatic Requirements Conforms None. 6.3.1 All VariousAllII.1 Various Departuredesign supports an exemption NuScale The Conforms None. 6.2.7 II.5II.6 Water Hammer Conforms None. 6.3.1 II.2II.3 Single-Failure Consideration Inservice Inspection and Operability Conforms None.II.8II.9 Long Term Cooling 6.3.1 Conforms None. 6.3.1 II.4 System Control Reactivity Combined Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 6.3, Rev 0: Emergency 6.3, Rev 0: Emergency DSRS Core Cooling System 6.3, Rev 0: Emergency DSRS Core Cooling System DSRS 6.2.6, Rev 0: DSRS Containment Leakage Testing SRP 6.2.7, Rev 1: Fracture SRP Prevention of Containment Pressure Boundary 6.3, Rev 0: Emergency DSRS Core Cooling System DSRS 6.3, Rev 0: Emergency 6.3, Rev 0: Emergency DSRS Core Cooling System 6.3, Rev 0: Emergency DSRS Core Cooling System DSRS 6.3, Rev 0: Emergency 6.3, Rev 0: Emergency DSRS Core Cooling System 6.3, Rev 0: Emergency DSRS Core Cooling System 6.3, Rev 0: Emergency DSRS Core Cooling System 6.3, Rev 0: Emergency DSRS Core Cooling System DSRS 6.3, Rev 0: Emergency 6.3, Rev 0: Emergency DSRS Core Cooling System
Tier 2 1.9-98 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.4 Not Applicable Not Applicable Not Applicable ts. This eliminates applicablecurrently only to Comments Section applicant responsibility. This guidance is applicable only to reactor for filtration emergency on rely that designs during a design habitability control room room basisNuScale control accident. The habitability system neither relies on nor uses emergencyprotect filtration to operators during accident conditions. Rather, clean air is provided using compressed air tanks. This guidance is applicable only to reactor room control the on rely that designs emergencycontrol system for ventilation room habitability during a design basis room accident. The NuScale control habitability system uses compressed air tanks as a clean air source during postulated accident even for radioactive material or the potential via the control room gases to enter toxic inlets. ventilation system applicable. is partially The subtier RG 1.183 6.4.1 operating reactors. Status Review Plan (SRP) and Design Specific Review Conforms Not Applicable Not Applicable Not Applicable This guidance is Standard (DSRS) (Continued) AC Title/Description Conformance Filtration System Control Room Operating Reactors That Do Not Implement an Alternative Source Term and Licensees That Implement an Term Alternative Source AC II.1II.2II.3 Control Room Emergency ZoneII.4 Ventilation Criteria System Pressurization Systems Conforms Emergency Standby Atmosphere None. Conforms None.II.5 Conforms None.and Relative Location of Source II.6.A 6.4 Dose Guidelines for Current II.6.B 6.4 II.7 6.4 Reactors Dose Guidelines for New Toxic Gas Hazards Partially Conforms Programmatic requirements are the COL Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 6.4, Rev 3: Control Room 6.4, Rev SRP System Habitability 3: Control Room 6.4, Rev SRP System Habitability 3: Control Room 6.4, Rev SRP System Habitability 3: Control Room 6.4, Rev SRP System Habitability 3: Control Room 6.4, Rev SRP System Habitability 3: Control Room 6.4, Rev SRP System Habitability 3: Control Room 6.4, Rev SRP System Habitability 3: Control Room 6.4, Rev SRP System Habitability
Tier 2 1.9-99 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.5.3 Not Applicable Not Applicable Comments Section use engineeredfilter use safety feature (ESF) systems or ESF ventilation systems to design basisa mitigate the consequences of accident. In the NuScale Power Plant design there is a nonsafety-related Reactor Building heating ventilatingand air conditioning system which includes filtering; however,in it is notthe credited dose analysis. This SRP section and its acceptance criteriasection and its acceptance SRP This to large are applicable only II.3) (II.1 through The with containment spray systems. LWRs NuScale containmentdoes vessel design incorporatenot a spray system. A portion of this acceptance criterion and its applicablesubtier guidanceis only to LWR designscontainment that include 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 removal systems such as containment heat serve a fission that containment spray) product removal/dose mitigation function. Rather, fission product control is inherent in Power the passive design of the NuScale wherein the compact containment Module, vessel is submerged in thereactor pool. Therefore, the aspects of this guidance related to theseare systemsnot applicable is applicable to This guidance the DCA. to containment of certain NuScale the review such as parameters and design features, designand leakage systems leakage rate isolation.prior to containment Status Review Plan (SRP) and Design Specific Review Not Applicable The NuScale Power Plant design does not Standard (DSRS) (Continued) AC Title/Description Conformance First full paragraph on Page 6.5.1-6, 6.5.1-6, on Page First full paragraph Design, Testing, and Maintenance of Air Cleanup System ESF Atmosphere and Adsorption Units Filtration AC II All VariousII.1 Primary Containment PartiallyConforms Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 6.5.1, Rev 4: ESF 6.5.1, Rev 4: ESF SRP Atmosphere Cleanup Systems 6.5.2, Rev 4: Containment SRP Spray as a Fission Product Cleanup System 6.5.3, Rev 3: Fission SRP Product Control Systemsand Structures
Tier 2 1.9-100 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Not Applicable Not Applicable Comments Section control system in the NuScalePlant Power the containment vessel in design is isolationcontainmentconjunction with the valves and passive containment isolation barriers. This acceptance criterion is applicable only This that incorporateLWRsto both a primary secondary containment.and NuScale The containment vessel design does not include a secondary containment. criteriasection and its acceptance SRP This are applicableonly to applicants for plant condenserdesigns that involve ice NuScalecontainments. The reactor design does not use an ice condenser containment. criteriasection and its acceptance SRP This that credit are applicableLWRs only to large a pressure suppression pool for fission retention (i.e., product scrubbing and BWRs). The NuScale reactor design does not credit or use a suppression pool. Status Review Plan (SRP) and Design Specific Review Not Applicable The only credited ESF fission product Standard (DSRS) (Continued) AC Title/Description Conformance Systems AC II.2 Secondary ContainmentII.4 Control Fission Product Other Not Applicable All VariousAll Various Not Applicable Not Applicable II.1II.2 Inspectionto Subject Components II.3 Conforms AccessibilityII.4 None. Examination Categories and Methods Conforms Inspection Intervals None. Conforms None. Conforms None. 6.6.1 6.6.3 6.6.2 6.6.4 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 6.5.3, Rev 3: Fission SRP Product Control Systemsand Structures 6.5.3, Rev 3: Fission SRP Product Control Systemsand Structures Rev 4: Ice 6.5.4, Draft SRP Condenser as a Fission Product Cleanup System 6.5.5, Rev 1: Pressure SRP Suppressionas Pool a Fission Product Cleanup System DSRS 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components
Tier 2 1.9-101 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.6.5 Table 6.6-1 Not Applicable Not Applicable Comments Section The operational program and by implementation milestones governed this acceptance criterion are the applicant.responsibilitythe COL of criteriasection and its acceptance SRP This are applicableonly to BWRs. Status Review Plan (SRP) and Design Specific Review Applicable None. Not Applicable Conforms None. 6.6.8 Standard (DSRS) (Continued) AC Title/Description Conformance Augmented ISI to Protect Against Protect ISI to Augmented Piping Failures Postulated AC II.9II.10 Code ExemptionsII.11 Requests Relief II.12 Code Cases Operational Programs Conforms None. Conforms None. Not Applicable Conforms None. 6.6 6.6 6.6 II.5II.6 Evaluation of ExaminationResults II.7 System Pressure Tests Conforms None. Supports Structural Conforms None. Conforms None. 6.6.5 All 6.6.7 Various 6.6.1 Not Applicable II.13 ProgramInformed ISI Risk Not II.8 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components DSRS 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components 3: Main 6.7, Draft Rev SRP Steam Isolation Valve Leakage Control System (BWR) DSRS 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components 6.6, Rev. 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components DSRS 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components DSRS 6.6, Rev 0: Inservice DSRS Testing of Inspection and Class 2 and 3 Components
Tier 2 1.9-102 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.1.1 6.2.2 6.3.2 ion is applicable but ion is applicable but n ofn this acceptance Comments Section This acceptance criter certain6-1, which would languagein BTP be applied by AcceptanceB.1, Criterion the NuScalerefersin to SSC that are not design.Specifically, the NuScale design spraycontainmentnotdoes system use a or operation, ECCS However, during a sump. collectNuScale inside theECCS water does containment vessel for recirculation back to the reactor core,and thus the intent of this applicableacceptancecriterion is to the DCA. of a portio The intent criterion is applicable but the specific language refers to that are not SSC in the NuScale design. Specifically the spraydesigncontainment a does not use system or a sump. However, during ECCS inside water does collect operation, ECCS for the NuScale containment vessel core, and to the reactor recirculation back containedthus the pH guideline in this applicableacceptancecriterion is to the DCA. This acceptance criter certain6-1, which would languagein BTP be applied by AcceptanceB.3, Criterion the NuScalerefersin to SSC that are not design.Specifically, the NuScale design spraycontainmentnotdoes system use a or operation, ECCS However, during a sump. collectNuScale inside theECCS water does containment vessel for recirculation back to the reactor core,and thus the intent of this applicableacceptancecriterion is to the DCA. Status Review Plan (SRP) and Design Specific Review Conforms Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Water Product Requirements for Fission Removal Aluminum Corrosion AC B.1pH for Emergency Coolant Minimum B.2 SpraypH and Water Chemistry Water B.3 Hydrogen Generation from Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: BTP 6-1, (March 2007): pH for (March 2007): pH for BTP 6-1, for Water Emergency Coolant Pressurized Water Reactors (March 2007): pH for BTP 6-1, for Water Emergency Coolant Pressurized Water Reactors (March 2007): pH for BTP 6-1, for Water Emergency Coolant Pressurized Water Reactors
Tier 2 1.9-103 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.0 Not Applicable Not Applicable Not Applicable Not Applicable able only to PWRs for Comments Section This guidance is applic which a postulatedLOCA results in core uncovery. For the NuScale design, a LOCA result in core uncovery. does not acceptanceThese criteria (B.1 through B.9) are applicableonly to large that LWRs incorporatebothsecondary a primary and containmentcontainment. The NuScale vessel design does not include a secondary containment. This guidance pertains to containment purge systems used to vent containment While the NuScale directly to the environs. containment vessel design includes an evacuation system, it serves a different purpose than a purge system, and includes features that provide suitable means to release to the environsprevent radiological NuScale (The II.13). 6.2.4, AC DSRS (see containment vessel evacuationsystem under valve closure times are addressed 6.2.4.) SRP Section This guidance is applicable only to LWR injection rely on safety that designs ECCS refueling (or borated)water pumps and storage tanks. The NuScale ECCS design water use pumps or refueling does not storage tanksequivalent). (or provides a general section DSRS This description of the process for reviewing I&C systems that is applicablethe DCA. to However, this guidance does not contain specific acceptanceSpecific criteria. 7 are criteria for SRP Chapter acceptance 7 provided in the individual SRP Chapter sections and are summarized in SRP Section Appendix 7.1-A. Table 7-1, and SRP SRP 7.1, Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance VariousVarious Not Applicable Not Applicable AC All (B.1 thru B.3) All VariousAll (B.1 thru B.5) Not Applicable All Various Not Applicable All Various Conforms Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: BTP 6-2, Rev 3: Minimum Rev BTP 6-2, Containment Pressure Model Performance ECCS for PWR BTP 6-3, Rev 3: Determination of Bypass Leakage Pathsin Dual Containment Plants 3: Containment Rev BTP 6-4, Normal Plant During Purging Operations the 3: Currently Rev BTP 6-5, Responsibility of Reactor Systems Piping From the and BWST) RWST (or Containment Sump(s)the to Pumps Safety Injection DSRS 7.0, Rev 0: DSRS Instrumentation and Controls Overview - Introduction and of Review Process
Tier 2 1.9-104 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.1 7.2 7.1.5 7.1.1 7.1.5 7.1.6 7.0.4 7.1.7 7.1.8 7.1.1 7.1.5 th the guidelines in Comments Section to the applicable regulatory guidance from the staff requirement the staff from guidance is SECY-93-087 memorandum to in Sectionsummarized 7.1.5. systems I&C in Section 7.1.6. summarized are designedthe quality assurance to in Sectionprogram described 17.5. None. 7.1.2 design is consistent wi design is NUREG/CR-6303. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms Conforms None. 7.1 7.1.2 Conforms of the NuScale I&C assessment The D3 Standard (DSRS) (Continued) AC Title/Description Conformance I&C - Hazard AnalysisI&C - Architecture System I&C - SimplicityI&C - Conforms ConformsSpecific SRP Acceptance Criteria None.Applicable to I&C Systems Important None. in Table 7.0-1. are Listed to Safety Ensure compliance to current version 1.75 RG of Ensure compliance to current version 1.152 RG of ConformsConformance with RG 1.53 None. Conforms None.SECY-93-087 7.1.8 7.0.3 ConformanceRG 1.53 to 7.1.6 ConformanceRG 1.62 to Conforms Conformance Conforms 7.1.3 Conforms None.RG 1.62 in Table See 1.9-2. 7.1.5 7.1.5 of reactor protection systems - - - - - Predictability and Repeatability Conforms None. 7.1.4 AC II.3II.4 GL 85-06 Conformsis 50.62 CFR Conformance to 10 II.2 All II.1 II.1II.2 analyses for performing D3 Methods II.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 7.1.3, Rev 0: DSRS Redundancy 7.1.5, Rev 0: Diversity DSRS and Defense in Depth 7.1.5, Rev 0: Diversity DSRS and Defense in Depth DSRS Appendix 7.0-A, Rev 0: Hazard Analysis I&C - DSRS Appendix 7.0-B, Rev 0: I&C - System Architecture DSRS Appendix 7.0-C, Rev 0: Simplicity I&C - DSRS 7.1.2, Rev 0: DSRS Independence DSRS 7.1.1, Rev 0: DSRS Fundamental Design Principles 7.1.2, Rev 0: DSRS Independence 7.1.4, Rev 0: DSRS and Predictability Repeatability 7.1.5, Rev 0: Diversity DSRS and Defense in Depth 7.1.5, Rev 0: Diversity DSRS and Defense in Depth 7.1.5, Rev 0: Diversity DSRS and Defense in Depth
Tier 2 1.9-105 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.1.2 7.1.5 7.2.2 7.2.3 Systems of Nuclear Nuclear of Systems Comments Section exceptions and and exceptions None. to the conform systems I&C safety Digital 7- 4.3.2- Std of IEEE 5.4 in Section guidance Digital Criteria for Standard IEEE 2003, Computers in Safety Power Generating Stations, as endorsed 7.1.1 (with identified 3. Rev. clarifications) by RG 1.152, to the conform systems I&C safety Digital and completion of reliability, integrity, protective action guidance in Sections 5.5, as Std 7-4.3.2-2003, IEEE 5.15 of and 3. Rev. endorsed by RG 1.152 Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance Conformance to IEEE Std. 7-4.3.2 Std. IEEE to Conformance ConformanceRG 1.28 to Conforms RG 1.152 to Conformance RG 1.168 to Conformance RG 1.169 to Conformance RG 1.170 to Conformance RG 1.171 to Conformance RG 1.172 to Conformance ConformsConforms Partially Conforms in TableRG 1.168 See RG 1.173 to Conformance 1.9-2.Conforms Partially in TableRG 1.169 See 4.3.2 Std 7- IEEE RG 1.28 in Table to See Conformance 1.9-2. 1.9-2.Conforms Partially in TableRG 1.152 See 1.9-2. in TableRG 1.170 See 1.9-2.Conforms Partially in TableRG 1.171 See 1.9-2.Conforms Partially in TableRG 1.172 See 1.9-2.Conforms Partially Conforms in TableRG 1.173 See 1.9-2.ConformanceRG 1.209 to 7.2.1 7.2.1 RG 1.151 to Conformance 7.2.1 7.2.1 7.2.1 7.2.1 RG 1.180 to Conformance 7.2.1 7.2.1 ConformsRG 1.204 to Conformance Conforms Partially in TableRG 1.151 See 1.9-2. None. Std 7-4.3.2IEEE to Conformance Conforms Partially in TableRG 1.180 See 1.9-2.Conforms Partially in TableRG 1.204 See 1.9-2. Conforms ConformanceRG 1.47 to 7.2.2 7.2.2 Conforms 7.2.2 None. 7.2.2 7.2.4 AC II.6 II.2 II.1 II.1 II.3 II.4 II.5 II.1 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 7.1.5, Rev 0: Diversity DSRS and Defense in Depth DSRS 7.2.1 Rev. 0: QualityDSRS II.1 7.2.2, Rev 0: Equipment DSRS Qualification 7.2.4, Rev 0: Operating DSRS and Maintenance Bypasses DSRS 7.2.1 Rev. 0: QualityDSRS 7.2.1 Rev. 0: QualityDSRS II.2 7.2.1 Rev. 0: QualityDSRS II.3 7.2.1 Rev. 0: QualityDSRS II.4 7.2.1 Rev. 0: QualityDSRS II.5 7.2.1 Rev. 0: QualityDSRS II.6 7.2.1 Rev. 0: QualityDSRS II.7 7.2.2, Rev 0: Equipment DSRS II.8 Qualification 7.2.2, Rev 0: Equipment DSRS Qualification 7.2.2, Rev 0: Equipment DSRS Qualification 7.2.2, Rev 0: Equipment DSRS Qualification 7.2.3, Rev 0: Reliability, DSRS Completion of and Integrity, Protective Action
Tier 2 1.9-106 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.2.7 7.2.5 7.2.6 7.2.8 7.2.9 7.2.7 7.2.10 formation associated ications) by RG 1.152 ications) by RG (with identified Comments Section ance of GL 91-04 is applicable to is 91-04 ance of GL ISA-67.04.01-2006, Setpoints Nuclear Safety-Related for Instrumentation, which addresses the issues 2006-17. identified in RIS For computer-based interlocks, the For computer-based andcomponents system conform to the IEEE Std 7- indigital computers guidance for endorsed as 4.3.2, exceptions and clarif Rev. 3. acceptance specific DSRS There are no criteria in this section. acceptance specific DSRS There are no criteria in this section. components and systems I&C safety Digital the identificationconform to guidance in Std 7-4.3.2-2003. 5.11 of IEEE Section acceptance specific DSRS There are no criteria in this section. However, the to review the guidance provided is used acceptability of the in with interaction between sense and commandother systems. features and the setpointasin described methodology NuScale Instrument TR-616-49121, Setpoint Methodology Technical(Ref. 7.2-27). Report Status Review Plan (SRP) and Design Specific Review Conforms The setpointconforms methodology to Standard (DSRS) (Continued) AC Title/Description Conformance ic Letter 91-04(GL) Conforms The guid Conformance to IEEE Std 7-4.3.2IEEE to Conformance Conforms RG 1.105 to Conformance (RIS) Issue Summary Regulatory NRC 2006-17 Conforms Partially in TableRG 1.105 See 1.9-2. Std 7-4.3.2IEEE to Conformance 7.2.7 ConformanceRG 1.75 to Conforms Conforms None.ConformanceRG 1.53 to ConformanceRG 1.62 to Conforms None. ConformsRG 1.62 in Table See 1.9-2. 7.2.9 7.2.12 7.2.11 AC All VariousII.1 Conforms II.2 II.1 All VariesII.1 Conforms All Various Conforms Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 7.2.7, Rev 0: SetpointsDSRS II.3 Gener DSRS 7.2.5, Rev 0: InterlocksDSRS II.1 DSRS 7.2.6, Rev 0: Derivation 7.2.6, Rev 0: Derivation DSRS Inputs System of 7.2.7, Rev 0: SetpointsDSRS II.1 7.2.9, Rev 0: Control of DSRS Access, Identification, and Repair 7.2.9, Rev 0: Control of DSRS Access, Identification, and Repair DSRS 7.2.11, Rev 0: Multi-Unit Stations DSRS 7.2.7, Rev 0: SetpointsDSRS II.2 DSRS 7.2.10, Rev 0: Interaction Between Sense and Command Features and Systems Other DSRS 7.2.12, Rev 0: Automatic and Manual Controls DSRS 7.2.8, Rev 0: Auxiliary 7.2.8, Rev 0: Auxiliary DSRS Features
Tier 2 1.9-107 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.4 8.1.4 8.2.2 8.3.1 8.3.2 8.2.3 7.2.14 7.2.15 7.2.13 7.2.13 7.2.15 Not Applicable Comments Section There are no specific DSRS acceptance specific DSRS There are no criteria in this section. However, the to review the guidance provided is used acceptability of information associated with human factors considerations. components and systems I&C safety Digital to the guidance related conform to in and calibration for test capability Std 7- IEEE of 5.5.2, and 5.5.3 Sections 5.7, 4.3.2-2003. in TableRG 1.118 See 1.9-2.NRC provides a matrix of the Table 8-1 DSRS guidance, and Commission requirements, policy documents, and industry codes and standardsareapplied that as acceptance of the andcriteria guidance to the review systems described in Sections 8.2, electrical Some of these and 8.4. 8.3.2, 8.3.1, 7.2.15 documents are not relevant or only partially relevant to the NuScale design. shutdown station are designed to maintain alarm system reliability in accordance with item II.T of SECY-93-087. responsibility of the COL applicant as described in Section 8.2.2. that includesGDC 17 from the associated for the offsite power system. requirements orms in TableRG 1.97 See 1.9-2. 7.2.13 Status Review Plan (SRP) and Design Specific Review Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance ConformanceRG 1.97 to ConformanceRG 1.47 to SECY-93-087 PartiallyConf Conforms None. Std 7-4.3.2IEEE to Conformance Conformsremote main control room and The Conforms ConformanceRG 1.118 to Specific SRP Acceptance Criteria Contained8.3.1, in SRP Sections 8.2, 8.4 (summarized in Table 8- 8.3.2, and PartiallyConforms 1) 7.2.4 Compliance5with GDC Compliance17 with GDC Not Applicable Conformance with GDC 5 is the Departuredesign supports an exemption NuScale The AC II.2 II.1 II.1 II.2 II (No Number) II.1 All Various Conforms II.2 II.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 7.2.13, Rev 0: Displays and Monitoring DSRS 7.2.13, Rev 0: Displays and Monitoring DSRS 8.1, Rev 0: Electric 8.1, Rev 0: Electric DSRS Power - Introduction DSRS 7.2.15, Rev 0: Capability andfor Test Calibration DSRS 7.2.15, Rev 0: Capability andfor Test Calibration DSRS 8.2, Rev 0: Offsite Power System DSRS 7.2.14, Rev 0: Human Factors Considerations DSRS 8.2, Rev 0: Offsite Power System DSRS 7.2.13, Rev 0: Displays and Monitoring
Tier 2 1.9-108 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.4 8.1 8.3 8.2.3 8.1.4 8.3.1 8.1.4 8.3.1 8.1.4 8.3.1 8.2.3 8.2.3 8.2.3 8.3.1 8.1.4 8.3.1 8.1.4 8.3.1 8.1.4 8.3.1 Not Applicable offsite power system. Comments Section The details regarding conformance with 10 CFR described in Section 8.4, are 50.63 Station Blackout. Development of the maintenance rule (10 CFR 50.65) program is theresponsibility the referencing the COL applicant of certified design. Onsite AC power systems conform to GDC 2 described in Section 8.3.1.2.1. the extent to Onsite AC power systems conform to GDC 4 described in Section 8.3.1.2.2. the extent to Onsite AC power systems conform to GDC 5 described in Section 8.3.1. the extent to from GDC 18 that includesGDC 18 from the associated requirements for the from GDC 33. GDC. these design criteria in lieu of that includesGDC 17 from the associated power for the onsite AC requirements system. that includesGDC 18 from the associated power for the onsite AC requirements system. from GDC 33. for electrical associated with GDC 50 penetration assemblies(EPAs) are included 8.3. in Section design criteria in lieu of these GDC. these design criteria in lieu of Status Review Plan (SRP) and Design Specific Review Conforms Departureprincipal complies NuScale with a set of Departureprincipal complies NuScale with a set of Standard (DSRS) (Continued) AC Title/Description Conformance Compliance18with GDC 41, 35, 38, with GDCs 34, Compliance 44 and Departurewith 10 Compliance CFR - 50.63 Passive Design design supports an exemption NuScale The Compliancewith 10 CFR 50.65(a)(4) Not Applicable with Compliance GDC 2 with Compliance GDC 4Compliance with GDC 5Compliance17with GDC Conforms Conforms Partially Conforms Compliance18with GDC Departuredesign supports an exemption NuScale The Departure50Compliancewith GDC design supports an exemption NuScale The Conforms The electrical design requirements and 44 and AC II.3 II.1 II.2 II.3 II.4 II.5 II.6 II.6 II.4II.4 II.5 Compliance33with GDC Departuredesign supports an exemption NuScale The II (No Number) Compliance33with GDC II (No Number) 41, 35, 38, with GDCs 34, Compliance Departuredesign supports an exemption NuScale The Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 8.2, Rev 0: Offsite Power System Power 8.3.1, Rev 0: AC DSRS Systems (Onsite) Power 8.3.1, Rev 0: AC DSRS Systems (Onsite) Power 8.3.1, Rev 0: AC DSRS Systems (Onsite) Power 8.3.1, Rev 0: AC DSRS Systems (Onsite) Power 8.3.1, Rev 0: AC DSRS Systems (Onsite) Power 8.3.1, Rev 0: AC DSRS Systems (Onsite) DSRS 8.2, Rev 0: Offsite Power System DSRS 8.2, Rev 0: Offsite Power System DSRS 8.2, Rev 0: Offsite Power System DSRS 8.2, Rev 0: Offsite Power System Power 8.3.1, Rev 0: AC DSRS Systems (Onsite) Power 8.3.1, Rev 0: AC DSRS Systems (Onsite)
Tier 2 1.9-109 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.1 8.3 8.3 8.3.2 8.3.2 8.3.2 8.3.2 8.3.2 Not Applicable Comments Section equipment is required to conform 10 CFR Std. 603-1991. 50.55a(h) and IEEE from GDC 17 that includesGDC 17 from the associated power for the onsite DC requirements systems. that includesGDC 18 from the associated power for the onsite DC requirements systems. from GDC 33. for electrical associated with GDC 50 penetration assemblies(EPAs) are included 8.3. in Section design. Development of the maintenance rule (10 CFR program including 50.65) the require that of SSC identification assessment per 10 CFR is the 50.65(a)(4) applicantresponsibilitythe COL of referencing the certified design. design criteria in lieu of these GDC. these design criteria in lieu of EDSS aspects of the applies to certain it As le No onsite electrical AC power system Status Review Plan (SRP) and Design Specific Review Departure Nuscale compliesseta of principal with Standard (DSRS) (Continued) AC Title/Description Conformance Compliancewith 10 CFR 50.65(a)(4) Not Applicable with 10 Compliance CFR 50.55a(h) Not Applicab with Compliance 10 CFR 52.47(b)(1) Conforms None. 8.1 and 44 and AC II.9 II.7 II.8 Number)II (No 2 GDC with Compliance Number)II (No 4 GDC with Compliance Number)II (No 5 GDC with Compliance II (No Number) Compliance17with GDC Conforms None. ConformsII (No Number) Compliance18with GDC None. Conforms Departure None.II (No Number) Compliance33with GDC design supports an exemption NuScale The II (No Number) 38, 41, 34, 35, with GDC Compliance DepartureII (No Number)50 Compliancewith GDC design supports an exemption NuScale The DepartureII.1design supports an exemption NuScale The 8.3.2 Conforms Conformance with RG 1.32 8.3.2 The electrical design requirements 8.3.2 PartiallyConforms Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 8.3.1, Rev 0: AC DSRS Power Systems (Onsite) DSRS 8.3.1, Rev 0: AC Power 8.3.1, Rev 0: AC DSRS Systems (Onsite) Power 8.3.1, Rev 0: AC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) DSRS 8.3.2, Rev 0: DC Power 8.3.2, Rev 0: DC DSRS Systems (Onsite)
Tier 2 1.9-110 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.4 19.3 8.3.2 8.3.2 8.3.2 8.3.2 8.3.2 8.3.2 8.4.3 8.3.2 8.3.2 Not Applicable ns staff visits to visits to ns staff Comments Section it applies to certain aspects of the EDSS aspects of the applies to certain it design. As it applies to certain aspects of the EDSS aspects of the applies to certain it As design. design. design. design. As describedAs in Section 8.4, all safety- related functions can be performed without for 72 hours after an on AC power reliance 19.3, a As described in Section SBO event. RTNSS process has beenimplemented. Consequently, the Alternate AC Power notSource is applicablethe to NuScale design. Althoughpower supplies DC are not mitigation SBO therequired to meet the 50.63, 10 CFR of requirements independenceof SBO related power is described in Section 8.3. supplies (EDSS) This SRP appendix gover plantpart sites as of licensing reviews stage.the operating or COL during CFR 50.65) program includingCFR 50.65) the require that of SSC identification is the 50.65(a)(4) CFR assessment per 10 applicantresponsibilitythe COL of referencing the certified design. EDSS aspects of the applies to certain it As EDSS aspects of the applies to certain it As EDSS aspects of the applies to certain it As nforms nforms in TableRG 1.63 See 1.9-2. 8.1 Status Review Plan (SRP) and Design Specific Review Partially Conforms None.Partially Conforms Partially Conforms 8.4 Standard (DSRS) (Continued) AC Title/Description Conformance Compliance with 10 Compliance CFR and 50.63 RGof the guidelines 1.155 Power Sources Alternate AC of Use of Passive for Plants RTNSS and Design powerIndependence of SBO-related sources AC II.2 Conformance with RG 1.75 Conforms As All Various Not Applicable II.3 Conformance with RG 1.81 PartiallyConforms II.4II.5II.6 with RG 1.118 Conformance II.7 with RG 1.153 Conformance Conforms Partially with RG 1.153 Conformance Conforms Partially Conformance with RG 1.63Conforms Partially PartiallyCo II.3 II.2 II.8 Conformance with RG 1.160 II.1 Not Applicable Development of the maintenance (10 rule Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 8.3.2, Rev 0: DC Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) 8-A, SRP Appendix Rev1: General Agenda, Station Site Visits DSRS 8.3.2, Rev 0: DC Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) DSRS 8.3.2, Rev 0: DC Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) DSRS 8.4, Rev 0: Station DSRS Blackout DSRS 8.4, Rev 0: Station DSRS Blackout DSRS 8.3.2, Rev 0: DC Power 8.3.2, Rev 0: DC DSRS Systems (Onsite) 8.4, Rev 0: Station DSRS Blackout
Tier 2 1.9-111 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.1.1 8.3.1 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable t apply to NuScale electric to NuScale t apply Comments Section e design, the offsite power power systems as these systems are not power systems as engineered safety and are notfeatures on to supportrelied engineered safety features. system doesto Class not supply power 1E loads and does not support safety-related functions. The NuScale design does not use safety The NuScale design does not use (or equivalent)injection tanks in response a designto basis accident. Design and NuScaleoperation of the ECCS alsonot do involve motor-operated valves. The backupdiesel generators are used only designated to power standby supplying for loads whenandneeded, are not interconnected with other AC power sources except for short periods for the purpose of load testing. The analysis of gridis the stability responsibility of the COL applicant that certification.referencesNuScale design the The analysis of gridis the stability responsibility of the COL applicant that certification.referencesNuScale design the disconnecting power to electrical componentsa fluid of system as one means designing against a singlefailureof that might cause an undesirable component action. Removal of electric powerfrom the in is not used safety-related valves designmeans as of satisfying a the single failure criterion. Status Review Plan (SRP) and Design Specific Review Conforms Not ApplicablenoBTP does This Not Applicable For the NuScal Standard (DSRS) (Continued) AC Title/Description Conformance Variousfor Purposes Diesel-Generator Sets Other Than SupplyingStandby Power is Prohibited Not Applicable VariousDesign Criteria Reflecting Importance of Providing Accurate Information to the Operator and Reducing the Not Applicable Adversely Affecting of Possibility BTP 8-4 establishes the acceptability of Monitored Safety Systems protectionfor the offsite power system to assure proper operation and sequencing of Class 1E loads AC All (B.1 thru B.4) B. Use of Onsite Emergency Power B.1B.2 Grid ReliabilityAll (B.1 through B.5) Grid Capacity Not Applicable All (B.1 thru B.6) Not Applicable Allvoltage Criteria for evaluating Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP BTP 8-1, Rev 3: Requirements on Motor- Operated Valves in the ECCS Accumulator Lines SRP BTP 8-2, Rev 3: Use of for Diesel-Generator Sets Peaking Rev 3: Stability of SRP BTP 8-3, Power Systems Offsite Rev 3: Stability of SRP BTP 8-3, Power Systems Offsite SRP BTP 8-4, Rev 3: SingleApplication of the Criterion to Manually Failure Controlled Electrically Operated Valves SRP BTP 8-5, Rev 3: Supplemental Guidance for Bypass and Inoperable Status Indication for Engineered Features Systems Safety SRP BTP 8-6, Rev 3: Adequacy of Station Electric Distribution System Voltages
Tier 2 1.9-112 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.1.2.1 9.1.2.3 9.1.2.3 9.1.2.3 9.1.2.3 9.1.1.1 9.1.2.3 9.1.2.5 Not Applicable Not Applicable em is not required is em Comments Section The NuScale plant does not require or The NuScale plant does not include safety-relateddiesel emergency generators. With the nonreliance on AC power for the operating functions, safety-related Technical Specifications restrictions (i.e., Allowed Outage Times) for inoperable AC power sources specified in this guidance are appropriatenot to apply. ESF ventilationAn syst (see RG 1.52). Status Review Plan (SRP) and Design Specific Review Not Applicable Not Applicable Not applicable to passivedesigns. plant Not Applicable Partially Conforms None.Partially Conforms None. 8.2.3 8.2.3 GDC 62 Conforms None. GDC 63 Conforms None. 9.1.1.3 9.1.2.1 Standard (DSRS) (Continued) AC Title/Description Conformance SpecificCriteria to Meet GDC 2SpecificCriteria to Meet GDC 4SpecificCriteria to Meet GDC 5SpecificCriteria to Meet ConformsGDC 61Criteria to Meet Specific None. ConformsSpecificCriteria to Meet None. Conforms Conforms None. 9.1.2.1 9.1.2.1 9.1.2.1 of of Providing Accurate Information to the Operator and Reducing the Adversely Affecting of Possibility Monitored Safety Systems open phase condition designs activeconditions plant for conditionspassive for plant designs AC All Design Criteria Reflecting Importance All VariousB.1B.2 Electrical system designaddress to B.3open phase Criteria for evaluating II.1 Not Applicable open phase Criteria for evaluating II.5 II.2 II.3 II.4 II.1 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP BTP 8-7, Rev 3: Criteria for Indications and Alarms Associated with Diesel- Generator Unit Bypassed and Inoperable Status SRP BTP 8-8, (Feb 2012): Onsite (Emergency Diesel Generators) and Offsite Allowed Sources Power Outage Time Extensions SRP BTP 8-9, Rev. 0: Open Electric Phase Conditions in Power System SRP BTP 8-9, Rev. 0: Open Electric Phase Conditions in Power System SRP BTP 8-9, Rev. 0: Open Electric Phase Conditions in Power System 9.1.1, Rev 3: Criticality SRP of Fresh and Spent Safety Fuel Storage and Handling DSRS 9.1.2, Rev 0: New and 9.1.2, Rev 0: New DSRS Spent Fuel Storage DSRS 9.1.2, Rev 0: New and 9.1.2, Rev 0: New DSRS Spent Fuel Storage and 9.1.2, Rev 0: New DSRS Spent Fuel Storage and 9.1.2, Rev 0: New DSRS Spent Fuel Storage DSRS 9.1.2, Rev 0: New and 9.1.2, Rev 0: New DSRS Spent Fuel Storage
Tier 2 1.9-113 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.1.3.1 9.1.3.2 9.1.3.3 9.1.3.1 9.1.3.3 9.1.2.2 9.1.2.3 9.1.3.3 9.1.3.2 9.1.3.5 ed accidents. The UHS except that: (1) The Comments Section normal makeup water supply system and its the I and seismic Category source are not systemnot is designed to QualityC Group sink (UHS) heat The ultimate 1.26. per RG supplysystem systemCategory I is a seismic and cooling fuel for spent source and shielding for accident conditions. A UHS makeup supply line is designed to Quality seismic Category I C and Group (2) An ESF ventilation system requirements. is not required (see RG 1.52). normal makeup water supply system and its to accommodate designed source are not postulatthe effects of andsystemthe supply is source system for shielding that are spent fuel cooling and designedaccommodate to the effects of postulated accidents. A UHS makeup supply ESF An 4. (2) GDC to meet line is designed ventilation system is not required (see RG 1.52). orms The design conforms except that: (1) orms This design conforms Status Review Plan (SRP) and Design Specific Review Conforms None. 9.1.2.1 Conforms None. 9.1.1 GDC 61 GDC 63 Conforms None. Conforms None. 9.1.3.1 9.1.3.1 Standard (DSRS) (Continued) AC Title/Description Conformance SpecificCriteria to Meet 10 CFR 20.1101(b) GDC 2SpecificCriteria to Meet Conf Partially GDC 4SpecificCriteria to Meet Conf Partially GDC 5SpecificCriteria to Meet SpecificCriteria to Meet ConformsSpecificCriteria to Meet None. 9.1.3.1 Margin AC II.2 II.3 II.6 II.7II.1 and Subcriticality Criticality Monitors II.4 II.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 9.1.3, Rev 0: Spent Fuel 9.1.3, Rev 0: Spent DSRS Pool Cooling and Cleanup System DSRS 9.1.3, Rev 0: Spent Fuel 9.1.3, Rev 0: Spent DSRS Pool Cooling and Cleanup System DSRS 9.1.2, Rev 0: New and 9.1.2, Rev 0: New DSRS Spent Fuel Storage DSRS 9.1.2, Rev 0: New and 9.1.2, Rev 0: New DSRS Spent Fuel Storage Fuel 9.1.3, Rev 0: Spent DSRS Pool Cooling and Cleanup System Fuel 9.1.3, Rev 0: Spent DSRS Pool Cooling and Cleanup System DSRS 9.1.3, Rev 0: Spent Fuel 9.1.3, Rev 0: Spent DSRS Pool Cooling and Cleanup System
Tier 2 1.9-114 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.1.3.3 Not Applicable Comments Section This acceptance criterion is applicable only This to COL applicants. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 9.1.3.1 14.3 Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance SpecificCriteria to Meet 10 CFR 20.1101(b) ITAAC forDesign Certification Applications ITAAC forCombined License Applications GDC 2SpecificCriteria to Meet GDC 5SpecificCriteria to Meet ConformsGDC 61Criteria to Meet Specific None. ConformsGDC 62Criteria to Meet Specific None. Conforms None.GDC 1SpecificCriteria to Meet Conforms None.GDC 2SpecificCriteria to Meet ConformsGDC 4SpecificCriteria to Meet None. ConformsGDC 5SpecificCriteria to Meet None. 9.1.4 Conforms None. 9.1.4 Conforms None. 9.1.4 9.1.4 9.1.5 9.1.5 9.1.5 9.1.5 AC II.8 II.6 II.1 II.2 II.3 II.4 II.1 II.2 II.3 II.4 II.7 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 9.1.3, Rev 0: Spent Fuel 9.1.3, Rev 0: Spent DSRS Pool Cooling and Cleanup System DSRS 9.1.3, Rev 0: Spent Fuel 9.1.3, Rev 0: Spent DSRS Pool Cooling and Cleanup System Load 9.1.4, Rev 4: Light SRP Handling System and Refueling Cavity Design Load 9.1.4, Rev 4: Light SRP Handling System and Refueling Cavity Design Load 9.1.4, Rev 4: Light SRP Handling System and Refueling Cavity Design Load 9.1.4, Rev 4: Light SRP Handling System and Refueling Cavity Design 9.1.5, Rev 1: Overhead SRP Handling Heavy Load Systems 9.1.5, Rev 1: Overhead SRP Handling Heavy Load Systems 9.1.5, Rev 1: Overhead SRP Handling Heavy Load Systems 9.1.5, Rev 1: Overhead SRP Handling Heavy Load Systems DSRS 9.1.3, Rev 0: Spent Fuel 9.1.3, Rev 0: Spent DSRS Pool Cooling and Cleanup System
Tier 2 1.9-115 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.2.7 9.2.7 9.2.7 (Used for Site for Site (Used for Site (Used for Site (Used Cooling Water Cooling Water Cooling Water Not Applicable System (SCWS)) System (SCWS)) System (SCWS)) System to the SCWS reviewed Comments Section The NuScale sitecooling water system essential cooling provide (SCWS) does not tosafety-related SSC and is not safety- related or important-to-safety. The applicability of GDC 2 under this acceptance criterion is limitedto aspectsa failureensuring of the that nonsafety-relatedin SCWS does not result on a Seismic Categoryan adverse effect I provided is design, this For the NuScale SSC. the of construction and the design by the to meet SCWS related nonsafety Regulatory Staff provisions of RG 1.29, C.1.i. Guidance system does site cooling water The NuScale not provide essential cooling to safety- and related SSC is not considered safety- related or risk-significant. The applicability NuScaleGDC 4 to thewater cooling of system reviewed under this acceptance criterion is limited to aspects ensuring that does SSC the nonsafety-related of a failure on a safety- in an adverse effect result not related SSC. system does site cooling water The NuScale not provide essential cooling to safety- or are not safety-related SSC and related risk-significant. The design and layout of 5. Specifically, satisfy GDC these systems sharing of thesite cooling water system between units has no reasonable likelihood of adversely affecting essential SSC and associated safety functions. The site cooling water system does not not does system water cooling site The serve a safety-relatedoraccident cooling mitigation function. Status Review Plan (SRP) and Design Specific Review Conforms Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Protection Against Natural Natural Protection Against 2) Phenomena (GDC Dynamic Effects Environmental and 4) (GDC Sharing of Structures, Systems, and (GDC 5) Components 44)Cooling Water System (GDC Not Applicable AC II.1 II.2 II.3 II.4 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 9.2.1, Rev 5: Station SRP Service Water System 9.2.1, Rev 5: Station SRP Service Water System 9.2.1, Rev 5: Station SRP Service Water System 9.2.1, Rev 5: Station SRP Service Water System
Tier 2 1.9-116 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.2.2 9.2.2 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Comments Section nction contemplated by this serve a safety-relatedoraccident cooling mitigation function. serve a safety-relatedoraccident cooling mitigation function. of environmentalis and dynamic effectsof provided in Sections 3.5 and 3.6. Criterion II.1. Criterion II.1. Criterion II.1. Criterion II.1. SRP criterion is applicable to the NuScale SRP criterion is applicable (RCCW) reactor component cooling water system. This criterion 1.29 is based on RG portions, and C.1 for safety-related position positionC.2 for nonsafety-related portions. notisPosition applicable C.1 sincethe NuScale The related. safety is not RCCW that the in with position C.2 complies RCCW affect failure could whose structural SSCs safety-related SSCs are the operability of designed as Seismic Category II. orms Additional information pertaining to impact Status Review Plan (SRP) and Design Specific Review Not ApplicableNot not does system water cooling site The ApplicableNot not does system water cooling site The Not Applicablecomment above See Acceptance for Partially ConformsPartially The system fu Standard (DSRS) (Continued) AC Title/Description Conformance Cooling Water System Inspection Inspection System Cooling Water 45) (GDC Cooling Water System Testing (GDC 46) Phenomena Components AC II.5 II.6 II.1 Natural Protection Against II.2II.3 EnvironmentalDynamic Effects and Conf Partially II.4 Sharing of Structures, Systems, and II.5 Cooling Water SystemII.6 Cooling Water System Inspection Not Applicable Cooling Water System Testing Not Applicablecomment above See Acceptance for comment above See Acceptance for Not Applicablecomment above See Acceptance for Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 9.2.1, Rev 5: Station SRP Service Water System 9.2.1, Rev 5: Station SRP Service Water System 9.2.2, Rev 4: Reactor SRP AuxiliaryWater Cooling System 9.2.2, Rev 4: Reactor SRP AuxiliaryWater Cooling System 9.2.2, Rev 4: Reactor SRP AuxiliaryWater Cooling System 9.2.2, Rev 4: Reactor SRP AuxiliaryWater Cooling System 9.2.2, Rev 4: Reactor SRP AuxiliaryWater Cooling System 9.2.2, Rev 4: Reactor SRP AuxiliaryWater Cooling System
Tier 2 1.9-117 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.2.4 9.2.5 9.2.5 RG 1.72). The NuScale Comments Section The NuScale potable and sanitary water systemswith do not interface systems potentially containing radioactivity. The NuScale potablesystems and sanitary water are designedwill not such that failure result other adverse impacts onor in flooding essential SSC. the is not applicable to Since RG 1.27 isNuScale design, compliance with GDC 2 demonstrated by adherence to RG 1.13, PositionsRegulatory C.1 and C.2.The spent fuel provides both NuScale UHS removal,andcooling and containment heat phenomenais protected from natural and site-related events by the Seismic Category I RXB structure and with a Seismic Category I emergency makeup line. acceptance criterion is applicable This except for aspects related to theof use fiberglass piping (see fiberglass piping.design does not use Status Review Plan (SRP) and Design Specific Review Conforms None. 9.2.5 Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Materials to the PWSW Phenomena Components AC II.1 ControlReleases of Radioactive of II.1 Natural Protection Against II.2II.3 Sharing of Structures, Systems, and II.4 Cooling Water SystemII.5 Cooling Water System Inspection PartiallyConforms Cooling Water System Testing Conforms None. Conforms None. 9.2.5 9.2.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 9.2.4, Rev 3: Potable and 9.2.4, Rev 3: Potable SRP Sanitary Water Systems 9.2.5, Rev 3: Ultimate SRP Heat Sink 9.2.5, Rev 3: Ultimate SRP Heat Sink 9.2.5, Rev 3: Ultimate SRP Heat Sink 9.2.5, Rev 3: Ultimate SRP Heat Sink 9.2.5, Rev 3: Ultimate SRP Heat Sink
Tier 2 1.9-118 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.2.6 9.2.6 9.2.8 10.4.7 10.4.7 10.4.7 10.4.7 10.4.7 Not Applicable actor Building. The Comments Section The NuScale design’s condensatestorageThe NuScale design’s systemneither is safety-related nor risk- systems storage condensate The significant. components are locatedand outside the Re Seismic Category I from a effects of discharging water condensate storage facility failure have no reasonable potentialto adversely impact or systems safety-related of operation the of the plant. safe operation Guidance Regulatory with Staff Consistent no portion of the NuScaleof RG 1.29, C.1.i condensate storage system requires design the safe- to withstand construction and shutdown earthquake to prevent a failure a Seismic affect adversely could that SSC. I Category the condensate storage facilities Sharing of of safety-related the ability impair not does or risk-significant SSC to perform their functions. safety The NuScale CHWS does not perform safety or containment isolation functions. This criterion is The CHWS based on RG 1.29. I. The Category Seismic as is not classified Regulatory Staff complies with CHWS whose failure that the SSC in C.1.i Guidance I Category Seismic affect adversely could II. Seismic Category are designed as SSC Status Review Plan (SRP) and Design Specific Review Conforms Conforms Conforms None.Conforms None. 9.2.6 9.2.6 Conforms Standard (DSRS) (Continued) AC Title/Description Conformance 10 CFR Compliance 20.1406 Conforms None. 9.2.6 Phenomena Phenomena Design Basis Design Environment Components AC II.1 Natural Protection Against II.1II.2 QualityRecords Standards and Natural Protection Against Not Applicable II.3Dynamic Effects and Environmental Conforms None. 9.2.8 II.2Dynamic Effects Environmental and II.4 Control of Radioactive Releases to the II.3 Sharing of Structures, Systems, and II.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 9.2.6, Rev 3: Condensate SRP Storage Facilities Water Rev 0: Chilled SRP 9.2.7, System Water Rev 0: Chilled SRP 9.2.7, System Water Rev 0: Chilled SRP 9.2.7, System SRP 9.2.6, Rev 3: Condensate SRP Storage Facilities 9.2.6, Rev 3: Condensate SRP Storage Facilities SRP 9.2.6, Rev 3: Condensate SRP Storage Facilities 9.2.6, Rev 3: Condensate SRP Storage Facilities
Tier 2 1.9-119 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.4 9.2.8 9.3.1 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Comments Section The NuScale CHWS is a nonsafety-system The NuScale CHWS functions. perform safety does not and is a nonsafety-system The NuScale CHWS functions. perform safety does not and is a nonsafety-system The NuScale CHWS functions. perform safety does not and is a nonsafety-system The NuScale CHWS functions. perform safety does not and pressure thanThe CHWS is at a higher the andsystems GRWS where theLRWS of introduction precluding interface, contaminants into the CHWS. radioactive non-areNuScale compressed air systems safety, non-risk-significant systems. non-areNuScale compressed air systems safety, non-risk-significant systems. andof this acceptance criterion The intent its subtier guidance - to maintain the ability to withstand and recover from a SBO lasting a specified minimum duration - are that refers applicable. However, language large as LWRs such designs plant reactor to the NuScale plant design. is not relevant to meets the intent The NuScale plant design of this guidance with its passive design and cope with power to reduced reliance on AC design-basis events. Specifically, to achieve not required compressed air is core cooling in the event of a station NuScaleblackout in the design. Status Review Plan (SRP) and Design Specific Review Not Applicable 10 CFR 50.63 Conforms Partially Standard (DSRS) (Continued) AC Title/Description Conformance Specific Criteria to Meet GDC 1Criteria to Meet Specific GDC 2Criteria to Meet Specific Not Applicable GDC 5SpecificCriteria to Meet Not Applicable SpecificCriteria to Meet Conforms None. 9.3.1 Components AC II.4II.5II.6 Sharing of Structures, Systems, and II.7 Cooling Water SystemII.8 Cooling Water System Inspection Cooling Water System Testing Not Applicable II.1 Minimization of Contamination Not Applicable II.2 Not Applicable II.3 Conforms II.4 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 9.2.7, Rev 0: Chilled Water Rev 0: Chilled SRP 9.2.7, System Water Rev 0: Chilled SRP 9.2.7, System Water Rev 0: Chilled SRP 9.2.7, System Water Rev 0: Chilled SRP 9.2.7, System Water Rev 0: Chilled SRP 9.2.7, System 9.3.1, Rev 2: Compressed SRP Air System 9.3.1, Rev 2: Compressed SRP Air System 9.3.1, Rev 2: Compressed SRP Air System 9.3.1, Rev 2: Compressed SRP Air System
Tier 2 1.9-120 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.3.2 9.3.3 9.3.4 9.3.4 Not Applicable e tank, pressurizer Comments Section CS safety-related function single-failurecriteria apply only to the This acceptance criterion is applicable This areBWR-specific,except for aspects that or (e.g., design NuScale part of the not refueling water storag relief tank, and containment sump). TSTF was addressed in NRC-approved This nois longer applicable. 366-A and portionsNo of the NuScale drain system the containment barrier. penetrate precludes inadvertentof the boron dilution reactor coolant system. two safety-relateddemineralized water isolationprovided valves preclude to an the reactor dilutioninadvertent boron of system. coolant Status Review Plan (SRP) and Design Specific Review Conforms None. 9.3.2 Conforms None.Conforms 9.3.3 Conforms None.Conforms None. 9.3.2 9.3.2 Partially ConformsPartially The only CV Standard (DSRS) (Continued) AC Title/Description Conformance 10 CFR Minimization 20.1406. of contamination ITAACCVCS Functional Performance during Adverse Environmental Phenomena; and defense-in-Pumping Capacity; depth RCS makeup Conforms None. 9.3.2 Phenomena Material to the Environment Design Classification AC II.3 Technical Specifications Not Applicable II.1 Sampling CapabilityConforms Partially II.1II.2II.3 Natural Protection Against II.1 Dynamic Effects and Environmental ControlReleases of Radioactive of ConformsII.2 None.GDC 5Failure Criteria and Single Conforms The 9.3.3 II.2 II.4II.5 Process Sampling System Functional II.6 Seismic Designand Quality Group Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 9.3.2, DSRS 9.3.2, Rev 3: Process and Sampling Post-Accident Systems DSRS 9.3.2, DSRS 9.3.2, Rev 3: Process and Sampling Post-Accident Systems 9.3.3, Rev 3: Equipment SRP System and Floor Drainage 9.3.3, Rev 3: Equipment SRP System and Floor Drainage 9.3.3, Rev 3: Equipment SRP System and Floor Drainage 9.3.4, Rev 0: Chemical DSRS Controland Volume System Boron (Including (PWR) System) Recovery 9.3.4, Rev 0: Chemical DSRS Controland Volume System Boron (Including (PWR) System) Recovery DSRS 9.3.2, DSRS 9.3.2, Rev 3: Process and Sampling Post-Accident Systems DSRS 9.3.2, Rev 3: Process and Sampling Post-Accident Systems DSRS 9.3.2, Rev 3: Process and Sampling Post-Accident Systems DSRS 9.3.2, Rev 3: Process and Sampling Post-Accident Systems
Tier 2 1.9-121 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 14.3 9.3.4 Not Applicable in subtier Bulletin Comments Section A portion of this acceptance criterionis of this acceptanceA portion but the specific languagerefers applicable relevant designs thatCVCS to are not to the design CVCS NuScale design. The does not have holdup tanks that are subject conditions the vacuum to this acceptancesentence80-05. The last of criterion is applicable to the NuScale CVCS appropriate include will which design, venting and drainingcapability. criteriasection and its acceptance SRP This are applicableonly to BWRs. Status Review Plan (SRP) and Design Specific Review Conforms is located outside the RCPB. The CVCS 9.3.4 Conforms None. 9.3.4 Conforms None. 9.3.4 Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Componentsquality of the RCPB, design seismic and classification classification Prevention of CVCS Holdup Tank Wall and Venting Buckling/Failure; CVCS Draining ITAACGDC 2 Conforms None. Conforms None. 9.3.4 9.3.6 Design and Arrangement Outside Containment AC II.3II.4 Minimization of contaminationII.5 Conforms None.II.6Volume Chemical Control System and II.7 Detection of Reactor Coolant Leakage II.8 9.3.4 All Various Not Applicable II.1 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 9.3.4, Rev 0: Chemical DSRS Controland Volume System Boron (Including (PWR) System) Recovery 9.3.4, Rev 0: Chemical DSRS Controland Volume System Boron (Including (PWR) System) Recovery 9.3.4, Rev 0: Chemical DSRS Controland Volume System Boron (Including (PWR) System) Recovery 9.3.4, Rev 0: Chemical DSRS Controland Volume System Boron (Including (PWR) System) Recovery 9.3.4, Rev 0: Chemical DSRS Controland Volume System Boron (Including (PWR) System) Recovery 9.3.4, Rev 0: Chemical DSRS Controland Volume System Boron (Including (PWR) System) Recovery 9.3.5, Rev 3: Standby SRP (BWR) System Control Liquid DSRS 9.3.6, Rev 0: DSRS Containment Evacuation and Flooding Systems
Tier 2 1.9-122 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.4.1 9.4.1 Comments Section plant operations. Up to12 Modules room. control same the share modules This acceptance criterion is applicable This ESF to for aspects related except The NuScale cleanup systems. atmosphere control room habitability system neither onreliesnor uses emergency filtration to protect operators during accident conditions. Rather, clean air is provided compressedusing air tanks. Status Review Plan (SRP) and Design Specific Review Conforms None.ConformsCRVS is part Operation of the of normal Conforms 9.4.1 Standard (DSRS) (Continued) AC Title/Description Conformance GDC 60TMI 10 CFR 50.34(f) Conforms Conforms None. None. 9.3.6 9.3.6 Phenomena Components Material to the Environment AC II.2 II.1II.2 Natural Protection Against II.3Dynamic Effects and Environmental II.4 Conforms Sharing of Structures, Systems, and II.5 None. Room Control ControlReleases of Radioactive of Conforms None. 9.4.1 9.4.1 II.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 9.3.6, Rev 0: DSRS Containment Evacuation and Flooding Systems 9.4.1, Rev 3: Control SRP Area Ventilation Room System 9.4.1, Rev 3: Control SRP Area Ventilation Room System 9.4.1, Rev 3: Control SRP Area Ventilation Room System 9.4.1, Rev 3: Control SRP Area Ventilation Room System 9.4.1, Rev 3: Control SRP Area Ventilation Room System DSRS 9.3.6, Rev 0: DSRS Containment Evacuation and Flooding Systems
Tier 2 1.9-123 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.4.1 9.4.2 stems (see RG 1.52). stems Comments Section The intent of this acceptance criterion andof this acceptance criterion The intent its subtier guidance - to maintain the ability withstandto recover from a station and blackout (SBO) lasting a specified minimum duration - is applicable. However, much of the specific language refers to reactor plant not is and LWRs, large as such designs plant design. The relevantthe NuScale to of design meets the intent NuScale plant thisits passive guidance design with and cope with power to reduced reliance on AC design basis events. Consistent with Commission policy, this coping capability eliminates safetytypical large LWR a benefit havinggains an alternate AC power by for source (e.g., gas turbine generator) station blackout. Moreover, and specific to Section 9.4.1 acceptance criterion, this SRP the control room habitability system (Section 6.4) relies on compressed air tanks pressurizeto the control room envelope in the event of an SBO. room and the main control The design of the surroundingwalls, ceiling, and structure maintainthe sink to as a passive heat act environment within acceptable conditions in the event of an SBO. acceptance criterion is applicable This ESF to for aspects related except cleanupatmosphere sy Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) g Current Power Conforms AC Title/Description Conformance Compliance with Compliance GDC 2 with Compliance GDC 5Compliance with GDC 60 Conforms None. Conforms Conforms None. 9.4.2 9.4.2 AC II.6 Loss of All Alternatin II.1 II.2 II.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 9.4.1, Rev 3: Control SRP Area Ventilation Room System Fuel 9.4.2, Rev 3: Spent SRP Ventilation System Pool Area Fuel 9.4.2, Rev 3: Spent SRP Ventilation System Pool Area Fuel 9.4.2, Rev 3: Spent SRP Ventilation System Pool Area
Tier 2 1.9-124 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.4.2 Not Applicable Not Applicable pplicable only to LWR Comments Section RWBV system does not filter exhaust. does RWBV system This acceptance criterion is applicable This ESF to for aspects related except described as systems, cleanup atmosphere in the comment above for RG 1.52 subtier to AcceptanceII.3. Criterion criteriasection and its acceptance SRP This (II.1 through II.3) are a designs that rely on the turbine area ventilation system, or portions thereof, to fulfill safety-related or risk-significant BuildingThe NuScale Turbine functions. to on is not relied (TBVS) system HVAC control airborne radioactivity concentrations in the Turbine Building and gaseous effluents during normal operations (including anticipated operational occurrences) and after accidentsthat result release.in a radioactivematerial Furthermore, there are no requirements for TBVS performance needed to preclude adverse effects on safety-related functions conditionsduring of plant operation. Exhaust is filtered by by the is filtered system. RBV Exhaust Status Review Plan (SRP) and Design Specific Review Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Compliance with GDC 61 with Compliance GDC 2 Conforms with Compliance GDC 5Compliance with GDC 60 Conforms None. 5, and 2, GDC with GDC Compliance ConformsGDC 60 Not Applicable None. The 9.4.3 9.4.3 AC II.4 II.1 II.2 II.3 thru All (II.1 II.3) Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 9.4.2, Rev 3: Spent Fuel 9.4.2, Rev 3: Spent SRP Ventilation System Pool Area SRP 9.4.3, Rev 3: Auxiliary and Area Ventilation Radwaste System SRP 9.4.3, Rev 3: Auxiliary and Area Ventilation Radwaste System SRP 9.4.3, Rev 3: Auxiliary and Area Ventilation Radwaste System 9.4.4, Rev 3: Turbine Area SRP Ventilation System
Tier 2 1.9-125 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.5.1 Not Applicable Not Applicable Not Applicable Not Applicable systems to mitigate systems to erion (RG 1.191) is erion (RG 1.191) criterion is applicable only Comments Section to reactor licenseesto seeking renewal. license This SRP Section addresses ESF ventilation systems designedfission for product in a post-design basis accident removal environment. The NuScale design does not on ESF ventilation rely of a design basis the consequences accident. Non-safety-related normal ventilation systems provide atmosphere cleanup capability,that meets as necessary, the design, testing, and maintenance These 1.140. specified in RG guidelines systems are not credited for meeting offsite dose limits. applicable Developmentrisk-implementation of a and informed, performance-based fire protectionresponsibility program is theof COL applicants that reference the NuScale design,and that elect to implement the provisions of 10 CFR 50.48(c). acceptance criterion is applicable This year will use the current except NuScale subtier documents. This acceptance crit applicableonly to reactor licensees that certifications the necessary have submitted for license termination under 10 CFR 50.82. Status Review Plan (SRP) and Design Specific Review Not Applicable Not Applicableacceptance This Not Applicable Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance NRC Staff Positions and Guidelines on Fire Protection Assessment (Including Appendix C) Considerations for License Renewal (Including Appendix B) Shutdown and Decommissioning Reactor Plants AC All VariousII.1 Not Applicable Fire Protection Probabilistic Risk II.2II.3 Fire Protection Program II.4 Fire ProtectionPermanently for Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 9.4.5, Rev 3: Engineered SRP Ventilation Feature Safety System 9.5.1.1, Rev 0: Fire SRP Protection Program 9.5.1.1, Rev 0: Fire SRP Protection Program 9.5.1.1, Rev 0: Fire SRP Protection Program 9.5.1.1, Rev 0: Fire SRP Protection Program
Tier 2 1.9-126 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.5.1 9.5.1 9.5.2 Table 9.5.1-2 Appendix 9A Appendix 9A Appendix Not Applicable Not Applicable to be applicableto le in all instances to to instances all le in Comments Section This acceptance criterion and itsacceptance subtier criterion and This guidance apply to COL applicantsunder 10 CFR 52. a COL applicants referencing certified design are responsible for implementing this guidance. Notwithstandingan the above, NuScale, as applicant for a design certification, considers this guidance applicationthe design certification to the the COL that ensure necessary to extent this guidance. applicant can satisfy The enhancedfor fire protection criteria new reactor designspassive specify separationof redundant trains as the preferred approach to ensure safe- shutdown capability. Due to the modular Powerthe NuScalesize nature and small of Module, it is not feasib provide installed passive separation of redundant trains. When train separationis redundant feasible, fire protection for not shutdown systems is employedensure, to practicable,onethe extent to such that shutdown divisionfire will be free of damage. acceptance criterion is the This applicant.responsibilitythe COL of Developmentrisk-implementation of a and informed, performance-based fire protectionresponsibility program is theof COL applicants that reference the NuScale design,and that elect to implement the provisions of 10 CFR 50.48(c). acceptance criterion is the This applicant.responsibilitythe COL of Status Review Plan (SRP) and Design Specific Review Not Applicable Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Reactor Combined License Applications (Including New Reactor Designs A) Appendix Implementation Milestones AC II.5 for New Fire Protection Program II.6Criteria for Fire Protection Enhanced II.7All OperationalProgram and Proposed Various Not Applicable II.1 Emergency Facilities and Equipment Partially Conforms Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 9.5.1.1, Rev 0: Fire SRP Protection Program 9.5.1.1, Rev 0: Fire SRP Protection Program 9.5.1.1, Rev 0: Fire SRP Protection Program 9.5.1.2, Rev 0: Risk- SRP Informed, Performance- Based Fire Protection Program DSRS 9.5.2, Rev 0: DSRS Systems Communication
Tier 2 1.9-127 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.5.2 9.5.2 9.5.2 Ch 14 Not Applicable this acceptance this acceptance Comments Section The NuScale standard plant designwill onsite technicalaninclude provisions for support centerand an onsite operational bysupport centeras specified 10 CFR 50.34(f)(2)(xxv) and this acceptance serving systems Communication criterion. theseemergency facilities in support of responseresponsibility are theCOL of the applicant. The NuScale design includes provisions for facilities (i.e., emergency design-specific design features, facilities, pertain to are that equipment and functions, technicallyto the NuScale standard relevant plant design), consistentwith 10 CFR and 50.47(a)(8) and systems Communication criterion. equipment serving these facilities in emergency response are thesupport of applicant.responsibilitythe COL of acceptancethisThe aspects criterion of within the scope of NuScale design are to related Aspects to the DCA. applicable site-specific design, fabrication, erection, construction, testing, and inspection of SSC, and maintenance of records for activities the are the facility, the life of throughout applicantresponsibilitythe COL of referencing the certified design. COL applicant responsibility to prepare COL-specific ITAAC. the COL applicant referencingcertified design. Status Review Plan (SRP) and Design Specific Review Not Applicable None. Not Applicable Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Emergency Facilities and Equipment for Meeting 10 CFR 52.47(a)(8) Design, Fabrication, Erection, Testing, and Inspection Construction, SSC to Meet 10of CFR 50.55a ITAACCOL applicantaITAAC for Compliance with GDC 1 Conforms Not Applicable Conforms Site-specific scope is the responsibility of Operational Support Center AC II.5 II.2 Onsite TechnicalCenter and Support II.6 II.3 II.4 II.7 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 9.5.2, Rev 0: DSRS Systems Communication DSRS 9.5.2, Rev 0: DSRS Systems Communication DSRS 9.5.2, Rev 0: DSRS Systems Communication DSRS 9.5.2, Rev 0: DSRS Systems Communication 9.5.2, Rev 0: DSRS Systems Communication 9.5.2, Rev 0: DSRS Systems Communication
Tier 2 1.9-128 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.5.2 9.5.2 Not Applicable , including intra- 10 CFR and the 73.45 Comments Section Design documents meet requirements of communication ensuring that for GDC-19 appropriateis provided at equipment locations inside the controlthe room with and all normal to support capability emergency operations plantand communications plant to emergencyand off-site facilities evencommunicationthe requirements in failure within a of a single event communicationsubsystem or the lossthe of normal power source. The design addresses communications so thatcontrol room communicationscontrol room can maintain with site and offsite entities during normal and accident conditions. acceptance criterion is applicable only This licensees subjectto to generalrequirements of performance 10design does not 73.20. The NuScale CFR use or transportor reprocess spent fuel special nuclear material. Site-specific, programmatic aspects of physical security communication systems are the responsibilityCOL of the applicant of Aspects referencing the certified design. this acceptance criterion related to the design of the powerphysical reactor and communication systems are within the of the certified design and are scope to the DCA. applicable Status Review Plan (SRP) and Design Specific Review Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Compliance with Compliance GDC 2 with Compliance GDC 3 with Compliance GDC 4Compliance with GDC 19 Conforms None. Conforms None. Conforms Conforms None.Compliance with 10 CFR 73.45(e)(2)(iii), 10 CFR and 73.45(g)(4)(i), 9.5.2 10 CFR 73.45(g)(4)(ii) 9.5.2 with Compliance 10 CFR 73.46(f) 9.5.2 Conforms AC II.8 II.12 II.13 II.9 II.10 II.11 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 9.5.2, Rev 0: DSRS Systems Communication DSRS 9.5.2, Rev 0: DSRS Systems Communication 9.5.2, Rev 0: DSRS Systems Communication DSRS 9.5.2, Rev 0: DSRS Systems Communication 9.5.2, Rev 0: DSRS Systems Communication 9.5.2, Rev 0: DSRS Systems Communication
Tier 2 1.9-129 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 13.6 Not Applicable Not Applicable 13.6 (via Security 13.6 (via Technical Report) Comments Section Site-specific, programmatic aspects of physical security communication systems are the responsibilityCOL of the applicant referencingdesign. Aspects of the NuScale arerelatedthis acceptance criterion that to the physical designof the powerreactor and communication systems within the of scope the certified design are applicable to the DCA. Design focus pertains to addressing for physicalrequirements protection of licensed activities in nuclear power reactors the against radiologicalsabotage and communicationnecessary requirements for of this designthis protection. Elements fall under the COL applicant and are addressed plan. security physical the facility as part of does notThe NuScale plant design require diesel or include safety-related emergency generators that are subject toSRP this power is relied upon or DC section. No AC for the performance of NuScale plant safety functions. does notThe NuScale plant design require diesel or include safety-related emergency generators that are subject toSRP this power is relied upon or DC section. No AC for the performance of NuScale plant safety functions. Status Review Plan (SRP) and Design Specific Review Conforms Not Applicable Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Compliance with 10 CFR 73.55(e)(9)(vi)(B) with Compliance 10 CFR 73.55(j) Conforms Partially 4, GDC 2, GDC with GDC Compliance 17 5, and GDC 4, GDC 2, GDC with GDC Compliance GDC 45, and GDC 44, 17, GDC 5, GDC 46 AC II.14 thru All (II.1 II.4) thru All (II.1 II.7) II.15 II.1II.2II.3the System Integrated Design of Emergency Lighting System(s) Conforms Lighting Levels None. Conforms None. Conforms None. 9.5.3 9.5.3 9.5.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 9.5.2, Rev 0: DSRS Systems Communication 9.5.4, Rev 3: Emergency SRP Diesel EngineFuel Oil Storage System Transfer and 9.5.5, Rev 3: Emergency SRP Diesel EngineCooling Water System DSRS 9.5.2, Rev 0: DSRS Systems Communication SRP 9.5.3, Rev 3: Lighting 9.5.3, Rev 3: Lighting SRP Systems 9.5.3, Rev 3: Lighting SRP Systems 9.5.3, Rev 3: Lighting SRP Systems
Tier 2 1.9-130 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Comments Section RS 10.2.3, Section I and DSRS 3.5.1.3, Section I and DSRS 3.5.1.3, RS 10.2.3, of barriers for theprotection of SSCs of effects of from the to safety important turbine missiles. of barriers for theprotection of SSCs of effects of from the to safety important turbine missiles. The NuScale plant design does notThe NuScale plant design require diesel or include safety-related emergency generators that are subject toSRP this power is relied upon or DC section. No AC for the performance of NuScale plant safety functions. does notThe NuScale plant design require diesel or include safety-related emergency generators that are subject toSRP this power is relied upon or DC section. No AC for the performance of NuScale plant safety functions. does notThe NuScale plant design require diesel or include safety-related emergency generators that are subject toSRP this power is relied upon or DC section. No AC for the performance of NuScale plant safety functions. There are no safety-related SSC in the Turbine Building. Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Status Review Plan (SRP) and Design Specific Review Not Applicable Not Applicable Not Applicable Not Applicablerelies on the use The NuScale plant design Not Applicable Not Applicablerelies on the use The NuScale plant design Standard (DSRS) (Continued) AC Title/Description Conformance Compliance with GDC 2, GDC 4, GDC 4, GDC 2, GDC with GDC Compliance 17 5, and GDC 4, GDC 2, GDC with GDC Compliance 17 5, and GDC 4, GDC 2, GDC with GDC Compliance 17 5, and GDC from to safety SSC important Protect athe effectsofmissiles with turbine turbine overspeed protection system 4) (GDC onPrevention of AdverseEffects TurbineSSC in theSafety-Related Building essential for overspeed protection. AC All (II.1 thru All (II.1 II.4) thru All (II.1 II.4) thru All (II.1 II.4) II.1 II.2II.3 valves Inservice Inspection covering II.1Selection Materials Not Applicable Per DS Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 9.5.6, Rev 3: Emergency 9.5.6, Rev 3: Emergency SRP Diesel Engine System Starting 9.5.7, Rev 3: Emergency SRP Lubrication Diesel Engine System 9.5.8, Rev 3: Emergency SRP Diesel Engine Combustion Air System Exhaust Intake and 10.2, Rev 3: Turbine SRP Generator 10.2, Rev 3: Turbine SRP Generator 10.2, Rev 3: Turbine SRP Generator DSRS 10.2.3, Rev 0: Turbine Rotor Integrity
Tier 2 1.9-131 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Comments Section 10.2.3, Section I and DSRS 3.5.1.3, Section I and DSRS 3.5.1.3, 10.2.3, DSRS 10.2.3, Section I and DSRS 3.5.1.3, Section I and DSRS 3.5.1.3, DSRS 10.2.3, Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Section I.1, plants that use barriers to protect essential SSCs specified in RG 1.115 ondo notmissile the turbine have to rely generation probabilities,turbine including rotor integrity. Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance 10 CFR 52.47(b)(1) ITAAC 10 CFR 52.47(b)(1) Not Applicable Section I and DSRS 3.5.1.3, DSRS 10.2.3, Per AC II.2 Fracture ToughnessII.3 Pre-Service Inspection Not ApplicableII.4 Per DSRS Turbine Rotor Design Not Applicable Section I and DSRS 3.5.1.3, DSRS 10.2.3, Per II.5 Inservice Inspection Not Applicable Section I and DSRS 3.5.1.3, DSRS 10.2.3, Per Not Applicable Per II.6 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 10.2.3, Rev 0: Turbine Rotor Integrity DSRS 10.2.3, Rev 0: Turbine Rotor Integrity DSRS 10.2.3, Rev 0: Turbine Rotor Integrity DSRS 10.2.3, Rev 0: Turbine Rotor Integrity DSRS 10.2.3, Rev 0: Turbine Rotor Integrity
Tier 2 1.9-132 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 10.3.1 10.3.1 10.3.1 Not Applicable Not Applicable t t as backup to the the applicability of is applicable. The Comments Section components. components. The NuScale main is not steam system (MSS) safety-related, but the portion of the system the main steam isolation downstream of the includes RXB valves (MSIV) inside the secondary which MSIVs ac by the is ensured Functionality MSIVs. design and constructionthe of MSS tothe Regulatory Staff provisions of RG 1.29, and C.2. C.1.i Guidance safety-related or is not The NuScale MSS risk-significant. Thus, to the NuScale MSS reviewed under GDC 4 to this acceptance criterion is limited aspectsa failureensuring of the that nonsafety-relatednotdoes result in an SSC adverse effect on a safety-related SSC. andof this acceptance criterion The intent its subtier guidance of design meets the intent NuScale plant thisits passive guidance design with and cope with power to reduced reliance on AC design basis events. Status Review Plan (SRP) and Design Specific Review Conforms Conforms Conforms None.Conforms None.Conforms 10.3.1 None. 10.3.1 10.4.1 Not Applicable The NuScale design contains no Class 2 or 3 Not Applicable The NuScale design contains no Class 2 or 3 Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Protection against natural phenomena (GDC 2) Protection of SSC important to safety missiles of turbine from the effects 4) (GDC to safety important SSC Shared functions safety required perform 5) (GDC MSS is capable of supporting core cooling or safe-shutdown (non-DBA) in the event of an SBO (10 CFR 50.63) Important-to-SafetyProtection of SSC from Tornado Missiles (RG 1.117, Positions 2 and 4) Appendix Prevent excessive releases of to the environment radioactivity 60) (GDC Class 2 and 3 Components Components AC II.4 II.1 II.1II.2II.1 Materials Selection and Fabrication of 3 and Class 2 of Toughness Fracture II.2 II.3 II.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 10.3, Rev 0: Main Steam DSRS Supply System DSRS 10.3, Rev 0: Main Steam DSRS Supply System and 10.3.6, Rev 3: Steam SRP Feedwater System Materials and 10.3.6, Rev 3: Steam SRP Feedwater System Materials 10.4.1, Rev 3: Main SRP Condensers DSRS 10.3, Rev 0: Main Steam DSRS Supply System 10.3, Rev 0: Main Steam DSRS Supply System 10.3, Rev 0: Main Steam DSRS Supply System
Tier 2 1.9-133 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 10.4.6 10.4.4 Comments Section he NuScale design supports an exemptionhe NuScale is outside the steam generator tubes, the generator tubes, steam is outside the isandthe tubes,secondary inside water nothereis SG blowdown so the secondary chemistry requirements for the NuScale in the outlined from those differ design referenced EPRI report. from the power provisionsthe powerfrom of GDC 34. As described in Sectionthe design 3.1.4, NuScale-specific principal complies with a design criterion in lieuGDC. of this BWR only.BWR only. Not Applicable Not Applicable Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 10.4.2 10.4.3 Conforms In the NuScale SG design, the primary water 10.4.5 Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Prevent excessive releases of to the environment radioactivity 60) (GDC Prevent excessive releases of to the environment radioactivity 60) (GDC 4)(GDC Failures Piping Residual Heat Removal (GDC 34) Departure Conforms Path Leakage Alternate MSIV T None. safety to important SSC of Flooding 4) (GDC watercycle BWR plant Maintain direct Applicable Not quality toavoid corrosion-induced coolantthe reactor failure pressure of boundary14) (GDC Maintain indirect cycle PWR water quality toavoid corrosion-induced coolantthe reactor failure pressure of boundary14) (GDC 2) (GDC Seismic Events 10.4.4 Conforms None. 10.4.7 AC II.1 II.1 II.1 II.1 II.2 II.3 II.1 II.1 II.2 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 10.4.7, Rev 0: Condensate and Feedwater System SRP 10.4.2, Rev 3: Main SRP Condenser Evacuation System 10.4.3, Rev 3: Turbine SRP Gland Seal 10.4.4, Rev 3: Turbine SRP Bypass System 10.4.4, Rev 3: Turbine SRP Bypass System 10.4.4, Rev 3: Turbine SRP Bypass System 10.4.5, Rev 3: Circulating SRP Water System Rev 3: Condensate SRP 10.4.6, Cleanup System Rev 3: Condensate SRP 10.4.6, Cleanup System
Tier 2 1.9-134 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 10.4.7 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable moval system (DHRS) (DHRS) system moval em used to transfer em used to Comments Section WS is notis WS a system used totransfer The intent of this acceptance criterion andof this acceptance criterion The intent 4 GDC satisfy - to guidance subtier its flow fluid SSC from to protecting related instability effects such as water hammer - is specific of theapplicable. However, much language in the subtier guidance refers to reactor plant designs such as large LWRs, and is not relevant to the NuScale plant design. generatorThe NuScale steam design does use a blowdownnot system. The NuScale design neither requires nor uses an auxiliary feedwater system. The NuScale decayheat re functions similar to an performs some auxiliaryHowever, feedwater system. compared to an auxiliary feedwater system, operation, in its design, differs DHRS the LOCA break small to the relationship and plant response. The CFWS is notis a syst The CFWS to an ultimate heat sink.heat to an ultimate heat sink.heat to an ultimate heat sink.heat Status Review Plan (SRP) and Design Specific Review Conforms None. 10.4.7 Standard (DSRS) (Continued) AC Title/Description Conformance Fluid Instabilities 4) (GDC Conforms Partially Sharing of Structures, Systems, and (GDC 5) Components 44) (GDC Capability Heat Removal Applicable Not Inspection (GDC 45)Testing (GDC 46) Not Applicable The CF Not Applicablenotis a system The CFWS usedtransfer to AC II.4 II.2 AllAll Various Various Not Applicable Not Applicable II.3 II.5 II.6 II.7Corrosion Accelerated Flow Conforms None. 10.4.7 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 10.4.7, Rev 0: Condensate and Feedwater System DSRS 10.4.7, Rev 0: Condensate and Feedwater System 10.4.8, Rev 3: Steam SRP Generator Blowdown System 10.4.9, Rev 3: Auxiliary SRP Feedwater System (PWR) DSRS 10.4.7, Rev 0: Condensate and Feedwater System DSRS 10.4.7, Rev 0: Condensate and Feedwater System DSRS 10.4.7, Rev 0: Condensate and Feedwater System DSRS 10.4.7, Rev 0: Condensate and Feedwater System
Tier 2 1.9-135 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Not Applicable Not Applicable Comments Section This guidance is applicable only to large PWRs that use Auxiliary Feedwater (AFW) system pumps poweredbyand electrical fulfillssteam sources. TheNuScale a DHRS functionAFW system at a similar as the DHRS design large PWR. The NuScale does operates via passive use pumps: it not natural circulation. does notThe NuScale plant design use a design. top-feed steam generator does notThe NuScale plant design use a preheat steam generatordesign. acceptance criterion is applicable only This to large PWRs that use a once-through design. The NuScale plantsteam generator as design does not involve an AFW system butat a typical foundbe large LWR, would fulfills a similar that does include the DHRS functiona typical as AFW system. However, the NuScale steam generator design precludes potential water hammer issues without providing DHRS water through an supply top discharge mounted externally as is prescribedheader acceptance by this criterion. Status Review Plan (SRP) and Design Specific Review Conforms None. 5.4.1 Not Applicable Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance RG 1.110RG 1.112Conforms Partially in TableRG 1.110 See 1.9-2.Conforms Partially in TableRG 1.112 See 1.9-2. 11.1 11.1 Feedwater System Pump Drive and Power Supply Diversity Top-Feed Steam Generator Designs Not Applicable Preheat Generator Designs Steam Not Applicable Supply Feedwater Auxiliary - Designs DesignsTestsTest and - Procedures AC All Design Guidelines for Auxiliary B.1 thru TFSGD B.4 B.1 thru PSGD B.4 OTSGD B.1 Once-ThroughGenerator Steam OTSGD B.2 Once-ThroughGenerator Steam II.1 II.2 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: BTP 10-1, Rev 3: Design RevBTP 10-1, Guidelines for Auxiliary Feedwater System Pump Drive and Power Supply Pressurized Diversity for Plants Water Reactor 4: Design RevBTP 10-2, Guidelines for Avoiding Water Hammers in Steam Generators 4: Design RevBTP 10-2, Guidelines for Avoiding Water Hammers in Steam Generators 4: Design RevBTP 10-2, Guidelines for Avoiding Water Hammers in Steam Generators 4: Design RevBTP 10-2, Guidelines for Avoiding Water Hammers in Steam Generators DSRS 11.1, Rev 0: Coolant DSRS Source Terms DSRS 11.1, Rev 0: Coolant DSRS Source Terms
Tier 2 1.9-136 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.1 11.1 11.1 11.3 Not Applicable pects of ANSI/ANS ement) Methodology Methodology ement) Methodology ement) Comments Section The NuScale design is similar to large PWRs large The NuScale design is similar to release effluent fleet for existing in the calculations. However, an alternate methodology is usedbecause the existing code was developed PWRGALE in the 1980s reactors of that the PWR of evaluation for NuScaletime and does not address the plant design. PWRs large The NuScale design is similar to release effluent fleet for existing in the calculations. However, an alternate methodology is usedbecause the existing code was developed PWRGALE in the 1980s reactors of the large PWR evaluation of for addressNuScale thedoes not that time and as Some design. plant coolant are used for the18.1 source terms. Decontamination factors are consistent Technical Report,the NuScale Effluent with Replac (GALE Release Results, Revision 0 (TR-1116-52065). and Decontamination factors are consistent Technical Report,the NuScale Effluent with Replac (GALE Release Results, Revision 0 (TR-1116-52065). and in TableRG 1.110 See 1.9-2. 11.1 Status Review Plan (SRP) and Design Specific Review Conforms None. 11.1 Conforms None. 11.2 Partially Conforms Partially Conforms Partially Conforms Partially Conforms Standard (DSRS) (Continued) nt releases to the AC Title/Description Conformance RG 1.140DC/COL-ISG-5normal operation and AOO sources of Conforms Partially in Table RG 1.140 gaseous and liquid radioactive 1.9-2. Not Applicable effluents Release rates should be developed that are consistent using methods or PWR-GALE86, NUREG-0017, with 18.1-1999 ANSI/ANS 11.1 RWMS system augmentations used in cost-benefit calculations are consistentthe guidance of RG with 1.110 reduce gaseous effluent releases to to releases effluent gaseous reduce the environment liquid efflue reduce environment in characterizing terms, used source liquid and gaseous effluents AC II.6 II.3 II.5 II.7II.8 Decontamination factors used to II.9 Decontamination factors applied to II.10 Primary and secondary coolant II.4 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 11.1, Rev 0: Coolant DSRS Source Terms DSRS 11.1, Rev 0: Coolant DSRS Source Terms DSRS 11.1, Rev 0: Coolant DSRS Source Terms DSRS 11.1, Rev 0: Coolant DSRS Source Terms 11.1, Rev 0: Coolant DSRS Source Terms 11.1, Rev 0: Coolant DSRS Source Terms 11.1, Rev 0: Coolant DSRS Source Terms DSRS 11.1, Rev 0: Coolant DSRS Source Terms
Tier 2 1.9-137 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.1 11.1 11.5 11.6 11.2.3 Comments Section The NuScale design is similar to large PWRs large The NuScale design is similar to release effluent fleet for existing in the calculations. However, an alternate methodology is usedbecause the existing code was developed PWRGALE in the 1980s reactors of the large PWR evaluation of for addressNuScale thedoes not that time and plant design. The design basis coolantfor source term partiallyfuel based on a failedNuScale is less than 0.25 much fraction percent, which is described in NuScale’s Technical Report, Release (GALE Replacement)Effluent Methodology and Results, Revision 0 (TR-1116-52065). acceptance criterion is applicable This arerelatedexcept to for aspects that performance of a site-specific cost-benefit analysis, which is the responsibilitythe of applicant. COL Status Review Plan (SRP) and Design Specific Review Conforms None. 11.1 Conforms None. 11.1 Conforms None.Conforms None.Conforms None. 11.2.2 11.2.2 11.2.2 Partially Conforms Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Objectives 10 CFR 10 50.34(b)(3), CFR 50.34a, and 10 CFR 52.79(a)(3). calculational technique or any source term parameter expected in reactor pool water and secondary coolant is based on a combinationof fuel fractions assumptions of failed Requirements Waste Management System Liquid Components Facilitate Operation and Maintenance AC II.1 Dose Design Meet to Capability II.11II.12 are activation products If neutron II.13 The designbasis coolant source term II.14 II.2II.3 Processing for Anticipated Design II.4 Housing Seismic Design of Structures II.5 Provisionsto Control Leakage and features control Automatic Conforms None. 11.2 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 11.2, Rev 0: Liquid 11.2, Rev 0: Liquid DSRS System Waste Management DSRS 11.1, Rev 0: Coolant DSRS Source Terms DSRS 11.1, Rev 0: Coolant DSRS Source Terms DSRS 11.1, Rev 0: Coolant DSRS Source Terms 11.1, Rev 0: Coolant DSRS Source Terms 11.2, Rev 0: Liquid DSRS System Waste Management 11.2, Rev 0: Liquid DSRS System Waste Management 11.2, Rev 0: Liquid DSRS System Waste Management 11.2, Rev 0: Liquid DSRS System Waste Management
Tier 2 1.9-138 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria II.8. 11.3.1 11.3.2 11.3.3 11.3.4 11.3.2 11.3.4 Not Applicable Not Applicable As listed above in As listed acceptance criteria II.1, II.1, II.3, II.4, II.5, and criterion is applicable only criterion is applicable only Comments Section to applicants for an early site permit. This acceptance criterion is applicable This arerelatedexcept to for aspects that performance of a site-specific cost-benefit analysis, which is the responsibilitythe of applicant. COL acceptance criterion is applicable This site-specificgovernexcept for aspects that responsibility of the activities that are the applicant. COL to applicants for an early site permit. II.1, II.3, II.4, II.5, and II.8. II.4, II.3, II.1, msacceptance criteria described above in As Status Review Plan (SRP) and Design Specific Review Conforms None. 11.3.1 Conforms None.Conforms None. 11.3.2 11.3.1 Conforms None.Conforms None. 11.3.2 11.3.3 Not Applicableacceptance This Not Applicableacceptance This Partially Conforms Partially Conforms Standard (DSRS) (Continued) g Gaseous Waste Gaseous g AC Title/Description Conformance Design, Testing, and Maintenance of Adsorbers Filters and Charcoal HEPA Applications Objectives Requirements and Components of Classification Housin Structures Facilitate Decommissioning, and Minimize Generation of Radwaste Management System Hydrogen Explosion Offgas or Waste Gas Storage Tank Charcoal Delay Bed Applications AC II.6II.7II.1 Exhaust ventilation system Criteria for Early Site Permit Dose Design Meet to Capability Conforms None. 11.3 II.2II.3 Processing for Anticipated Design II.4 Seismic Designand Quality Group II.5 Minimize Contamination, Features to II.6II.7II.8 features control Automatic of Effects to Withstand Design II.9 Postulated Leakage or Failure of a II.10 Conforms Criteria for Early Site Permit None. RelevantBTP RGs, ISG, and PartiallyConfor 11.3.7 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 11.2, Rev 0: Liquid 11.2, Rev 0: Liquid DSRS System Waste Management DSRS 11.2, Rev 0: Liquid 11.2, Rev 0: Liquid DSRS System Waste Management 11.3, Rev 0: Gaseous DSRS System Waste Management DSRS 11.3, Rev 0: Gaseous 11.3, Rev 0: Gaseous DSRS System Waste Management 11.3, Rev 0: Gaseous DSRS System Waste Management 11.3, Rev 0: Gaseous DSRS System Waste Management 11.3, Rev 0: Gaseous DSRS System Waste Management DSRS 11.3, Rev 0: Gaseous 11.3, Rev 0: Gaseous DSRS System Waste Management 11.3, Rev 0: Gaseous DSRS System Waste Management 11.3, Rev 0: Gaseous DSRS System Waste Management 11.3, Rev 0: Gaseous DSRS System Waste Management 11.3, Rev 0: Gaseous DSRS System Waste Management
Tier 2 1.9-139 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 12.3 11.4.2 11.4.1 11.4.3 11.4.1 11.4.2 11.4.1 11.4.2 11.4.1 11.4.2 Table 11.4-9 Table 11.4-1 Not Applicable Not Applicable Table 11.4-5 thru Table 11.4-5 a Process Control Comments Section This acceptance criterion is applicable This except for aspects related to development and implementation of Program (PCP),which is the responsibility of the COL applicant. The development and implementation of a responsibilityPCP is theof the COL applicant. The development and implementation of a arethe responsibilityODCMPCP and of the applicant. COL acceptance criterion is applicable This site-specificgovernexcept for aspects that responsibility of the activities that are the applicant. COL certificationexcept for site-specific, programmatic and operational aspects that are the responsibilityCOL of the applicant. certificationexcept for site-specific, programmatic and operational aspects that are the responsibilityCOL of the applicant. orms This guidance is applicable to design Status Review Plan (SRP) and Design Specific Review Conforms None. 11.4-1 Table Conforms None. 3.8 Not Applicable Not Applicable Partially Conforms Partially Conforms This guidance is applicable to design Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Accordancewith Process Control Program Expected Radionuclide Distributions and Concentrations Waste in Accordance with Process Control Program Treatment Methods, Expected and Releases, Monitoring Effluent Control Instrumentation Setpoints Packaging Waste and Classification of Components and and Components of Classification Structures Housing Solid Waste Management System Facilitate Operation and Maintenance AC II.2II.3 SizingEquipment of Processing and Liquid Wet Waste Stabilization in Conforms None. 11.4.2 II.1 Design Parameters Based on II.4II.5 Wet Stabilizationof Forms of Other II.6 Design Objectives, Design Criteria, WasteShipping Containers, Casks, II.7II.8 Onsite Waste Storage FacilitiesConf Partially Seismic Designand Quality Group II.9 Provisionsto Control Leakage and Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 11.4, Rev 0: Solid Waste 11.4, Rev 0: Solid DSRS Management System Waste 11.4, Rev 0: Solid DSRS Management System DSRS 11.4, Rev 0: Solid Waste 11.4, Rev 0: Solid DSRS Management System DSRS 11.4, Rev 0: Solid Waste 11.4, Rev 0: Solid DSRS Management System Waste 11.4, Rev 0: Solid DSRS Management System Waste 11.4, Rev 0: Solid DSRS Management System DSRS 11.4, Rev 0: Solid Waste 11.4, Rev 0: Solid DSRS Management System Waste 11.4, Rev 0: Solid DSRS Management System DSRS 11.4, Rev 0: Solid Waste 11.4, Rev 0: Solid DSRS Management System
Tier 2 1.9-140 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 12.3 11.4.1 11.4.3 11.4.2 11.4.3 11.4.1 11.4.2 11.4.2 11.4.2 Not Applicable ant responsibility. Comments Section This acceptance criterion is applicable This site-specificgovernexcept for aspects that responsibility of the activities that are the applicant. COL no long term The NuScale design has storage facility for solid radioactive waste. is a COL applic This acceptanceThis criterion governs site- the PCP specific, programmatic aspects of that are development and implementation the responsibilityCOL of the applicant. This the extent guidance is applicable to necessary to ensure that theCOL applicant referencingcan satisfy the certified design the guidance. acceptanceThis criterion governs site- the PCP specific, programmatic aspects of that are development and implementation the responsibilityCOL of the applicant. This the extent guidance is applicable to necessary to ensure that theCOL applicant referencingcan satisfy the certified design the guidance. acceptanceThis criterion governs site- PCP specific, programmatic aspects of implementation (specific to mixed waste processing and disposal) that are the responsibility of the COL applicant. This the extent guidance is applicable to necessary to ensure that theCOL applicant referencingcan satisfy the certified design the guidancecontained therein. acceptance criterion is applicable This specific, programmatic except for site aspectsthat are the responsibility of the applicant. COL Status Review Plan (SRP) and Design Specific Review Not Applicable Partially Conforms Partially Conforms Partially Conforms Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance All Effluent Releases Associated with Releases Associated All Effluent Operation of the SWMS Facilitate Decommissioning, and Minimize Generation of Radwaste Onsite Storage (Including Appendix 11.4A) Disposing of Liquid,Dry Wet, and Solid Wastes Disposing of Liquid,Dry Wet, and Solid Wastes Wastes AC II.10Minimize Contamination, Features to II.11II.12Term for Long Design Storage Facility and - Processing A, B, C Class II.13 Greater thanC - Processing Class and II.14 ProcessingDisposing and of Mixed II.15 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 11.4, Rev 0: Solid Waste 11.4, Rev 0: Solid DSRS Management System DSRS 11.4, Rev 0: Solid Waste 11.4, Rev 0: Solid DSRS Management System Waste 11.4, Rev 0: Solid DSRS Management System Waste 11.4, Rev 0: Solid DSRS Management System Waste 11.4, Rev 0: Solid DSRS Management System Waste 11.4, Rev 0: Solid DSRS Management System
Tier 2 1.9-141 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.2 11.3 11.5 11.5 11.5 12.3 11.5 11.5 9.3.2 9.3.2 9.3.2 12.3.4 Not Applicable Comments Section ce criterion is applicable The information governed by this acceptance criterion is site-specific and is the responsibilityCOL of the applicant. acceptance criterion is applicable This except for certainaspectssubtier of its 1.97, RG RG 1.33, RG RG 1.21, (see guidance 7-10). 4.15, and BTP RG 4.1, acceptance criterion is applicable This except for certainaspectssubtier of its 1.97, RG RG 1.33, RG RG 1.21, (see guidance 7-10). Administrative and BTP 4.21 RG 4.15, procedural controlsand applicant are COL responsibility. acceptance criterion is applicable This except for certainaspectssubtier of its (see guidance RG 1.21, 1.33, RG 1.97 and Administrative and procedural RG 4.15). controls are COL applicant responsibility. acceptance criterion is applicable This except for certainaspectssubtier of its BTP 7-10). and RG 1.97 (see guidance Administrative and procedural controls are COL applicant responsibility. except for the administrative and that are the COL procedural controls applicant's responsibility. Status Review Plan (SRP) and Design Specific Review Conforms None. 11.4.1 Partially Conforms Partially Conforms Partially Conforms Partially Conforms Partially Conforms This acceptan Standard (DSRS) (Continued) AC Title/Description Conformance Monitoring Equipment and Sampling andPotential of Normal Analyses and Pathways Effluent Seismic design of structures housing SWMS Equipment and Sampling and Sampling and and Equipment Analysis of Radioactive Waste Process Systems (Including Appendix 11.5A) Procedural(Including Controls Appendix 11.5A) Gaseous of All Identified Effluent Release Paths(Including Appendix 11.5A) Release Effluent Liquid All Identified Paths AC II.1 Installation of Instrumentation and II.16 Operational Programs Not Applicable II.17II.18 Automatic control featuresII.19 Design of exhaust ventilation systems Conforms Not Applicable None.II.2 Instrumentation and Monitoring II.3for Administrative and Provisions II.4 Monitoring, Sampling, and Analyses 11.4.2 Not Applicable II.5 Monitoring, Sampling, and Analysis of Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 11.5, Rev 0: Process and Radiological Effluent Monitoring Instrumentation Systems and Sampling DSRS 11.4, Rev 0: Solid Waste 11.4, Rev 0: Solid DSRS Management System DSRS 11.4, Rev 0: Solid Waste 11.4, Rev 0: Solid DSRS Management System Waste 11.4, Rev 0: Solid DSRS Management System Waste 11.4, Rev 0: Solid DSRS Management System DSRS 11.5, Rev 0: Process and Radiological Effluent Monitoring Instrumentation Systems and Sampling DSRS 11.5, Rev 0: Process and Radiological Effluent Monitoring Instrumentation Systems and Sampling DSRS 11.5, Rev 0: Process and Radiological Effluent Monitoring Instrumentation Systems and Sampling DSRS 11.5, Rev 0: Process and Radiological Effluent Monitoring Instrumentation Systems and Sampling
Tier 2 1.9-142 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.5 11.6 11.6 11.6 12.3 9.3.2 Not Applicable Comments Section The information governed by this acceptance criterion is site-specific and is the responsibilityCOL of the applicant. acceptance criterion is applicable This except for aspects of its subtier regulation 10 CFR address testing 50.34(f)(2)(xxvi) that operationalprograms,and which are a COL applicant responsibility. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 11.5 11.5 Conforms None. 11.5 Conforms None. 11.6 Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Compliance with GDC 63 & 64 via 63 & with GDC Compliance items plan action post-TMI instrumentation used in primary and secondary coolant system leakage detection sampling equipment should be monitored airborne concentrationmonitoring locationsand sampling points AC II.6 Operational Programs Not Applicable II.7II.1 Descriptions of designfeatures and or Installation of instrumentation II.2 Gaseous and liquid release points II.3 Radiation exposurerates and II.4 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 11.5, Rev 0: Process and Radiological Effluent Monitoring Instrumentation Systems and Sampling DSRS 11.5, Rev 0: Process and Radiological Effluent Monitoring Instrumentation Systems and Sampling 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring
Tier 2 1.9-143 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.6 11.6 11.5 11.6 11.6 9.3.4 9.3.6 Ch 17 12.3.6 Not Applicable Comments Section ce criterion is applicable The RETS/SREC and ODCM are COL are COL The and ODCM RETS/SREC applicant responsibilities. except for site-specific, programmatic aspects regarding software reviews, which are the COL applicant's responsibility. Status Review Plan (SRP) and Design Specific Review Conforms None. 5.2.5 Not Applicable Partially Conforms Partially in TableRG 4.21 See 1.9-2. 11.5 Partially Conforms This acceptan Standard (DSRS) (Continued) used to develop,used to AC Title/Description Conformance RETS/SREC and ODCM established ODCM and RETS/SREC setpoints. with 10 Compliance CFR via 20.1406 and 07-07. 97-06, 08-08A NEI RG 4.21, instrumentation used in primary and secondary coolant system leakage detection review, verify, validate and audit digital computer software. AC II.5 Ensure samples are representative Conforms None. 9.3.2 II.8 II.9 and design features Description of II.6 Describe process II.7 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring
Tier 2 1.9-144 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 14.2 14.3 11.6 12.3 11.6 12.3 11.5 11.6 Ch 16 11.4.1 11.4.2 12.3.4 pplicable except for Comments Section This guidance is a aspects related to PCP development and are applicable to COL implementation that applicants. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 13.4 11.5 Conforms None. 11.5 Standard (DSRS) (Continued) AC Title/Description Conformance radiation monitoring equipment Radiation monitoring and sampling Initial conformance to Tech Specs, and ITAAC. Test Program, experience AC II.12 of ranges types and Describe the II.10information Additional on operating B.1 Processing Requirements Conforms II.13storage Reactor fuel monitors area Conforms None. 7.2 II.11 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring 4: Design RevBTP 11-3, Solid Guidance for Radioactive Waste Management Systems Light-Water- in Installed Cooled Nuclear Power Reactor Plants DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring DSRS 11.6, Rev 0: Guidance 11.6, Rev 0: Guidance DSRS on Instrumentation and Features for Control Design Process and Effluent Radiological Monitoring, and Airborne and Area Radiation Radioactivity Monitoring
Tier 2 1.9-145 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.4.1 11.4.2 11.4.2 11.3.1 11.3.3 11.4.2 Not Applicable pplicable except for pplicable except for Comments Section This guidance isguidance related to PCP This that are development and implementation applicableto COL applicants. This guidance is a use of and control to related aspects portable solid radwaste processing to COL that are applicable equipment applicants. This guidance is a aspects related to PCP development and are applicable to COL implementation that applicants. acceptance criterion is applicable This areBWR-specificexcept for aspects that or are site-specific. Status Review Plan (SRP) and Design Specific Review Not Applicable Partially Conforms Standard (DSRS) (Continued) atures Conforms Partially AC Title/Description Conformance or Dewatering Analysis AC B.2 Assurance of Complete Stabilization B.3 Waste StorageB.4Waste Systems Portable Solid B.5 ConformsConforms Partially None. Additional Design Fe B.1B.2 Waste Gas System Leak or Failure for AnalysisMethod Staff 11.4.1 Conforms None. 11.3.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: BTP 11-3, Rev 4: Design RevBTP 11-3, Solid Guidance for Radioactive Waste Management Systems Light-Water- in Installed Cooled Nuclear Power Reactor Plants 4: Design RevBTP 11-3, Solid Guidance for Radioactive Waste Management Systems Light-Water- in Installed Cooled Nuclear Power Reactor Plants 4: Design RevBTP 11-3, Solid Guidance for Radioactive Waste Management Systems Light-Water- in Installed Cooled Nuclear Power Reactor Plants 4: Design RevBTP 11-3, Solid Guidance for Radioactive Waste Management Systems Light-Water- in Installed Cooled Nuclear Power Reactor Plants 4: Postulated Rev BTP 11-5, Radioactive Releases Due to a Leak or Waste Gas System Failure 4: Postulated Rev BTP 11-5, Radioactive Releases Due to a Leak or Waste Gas System Failure
Tier 2 1.9-146 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.2.3 12.1.1 Not Applicable Not Applicable Not Applicable Not Applicable is site-specific is and Comments Section parameters under this acceptance criterionparameters under this acceptance (and SRP Section 2.4.13) applicableto COL applicant. criterionparameters under this acceptance is site-specificand applicable to COL applicant. applicant.responsibilitythe COL of This acceptance criterion is applicable This areBWR-specificexcept for aspects that or are to site-specific activities that are related the responsibilityCOL of the applicant. These site-specific aspects are the applicantresponsibilitythe COL of referencing the certified design. responsibility of the COL applicant.responsibilitythe COL of icableacceptance criterion is the This pplicable COL applicant. Not Applicable Status Review Plan (SRP) and Design Specific Review Not Applicableapplicant. COL Not Applicable Not Applicable The development of representative site Not Applicable The development of representative site Not Applicable Compliance with this guidance is the Standard (DSRS) (Continued) AC Title/Description Conformance Releases in Groundwater or Surface Water Criteria Radioactivity Concentration Levels AC B.1B.2 Failure Mechanism and Radioactivity B.3 Mitigating Design FeaturesB.4 Radioactive Source Term Not A B.5 Calculations of Transport Capabilities Partially conforms B.6 Exposure Scenariosand Acceptance B.7 SRP Dose Acceptance CriteriaII.1 Not Appl Specifications on Tank Waste II.2 Policy Considerations Design ConsiderationsConforms Partially Conforms None. 12.1.2 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: BTP 11-6, Rev 4: Postulated 4: Postulated Rev BTP 11-6, Radioactive Releases Due to Liquid-Containing Tank Failures 4: Postulated Rev. BTP 11-6, Radioactive Releases Due to Liquid-Containing Tank Failures 4: Postulated Rev BTP 11-6, Radioactive Releases Due to Liquid-Containing Tank Failures 4: Postulated Rev BTP 11-6, Radioactive Releases Due to Liquid-Containing Tank Failures (Rev 4): Postulated BTP 11-6, Radioactive Releases Due to Liquid-Containing Tank Failures SRP 12.1, Rev Assuring That 4: Occupational Radiation As Low Is Exposures Are Reasonably Achievable SRP 12.1, Rev Assuring That 4: Occupational Radiation As Low Is Exposures Are Reasonably Achievable BTP 11-6, Rev 4: Postulated 4: Postulated Rev BTP 11-6, Radioactive Releases Due to Liquid-Containing Tank Failures 4: Postulated Rev BTP 11-6, Radioactive Releases Due to Liquid-Containing Tank Failures
Tier 2 1.9-147 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.3 11.4 11.6 12.4 Ch 3 Not Applicable Not Applicable Not Applicable Not Applicable Comments Section Criterion II.3. This guidance governs site-specific operational programs, plans, and procedures that are the responsibility of the applicant. COL in TableRG 1.7 See 1.9-2. radiationnofrom source There is created of gaseousthe determination in containmentconcentrationsan following accident (such as sampleoutside lines containment). in TableRG 1.7 See 1.9-2. radiationno field created There is the from determination of gaseous concentrations in containment following an accident (such as sample lines outside containment). Applicablecomment above See Acceptance for Status Review Plan (SRP) and Design Specific Review Conforms None. 12.2 Conforms None. 12.3 Standard (DSRS) (Continued) AC Title/Description Conformance RG 1.183RG 1.7RG 1.112Conforms Partially in TableRG 1.183 See 1.9-2. 18.1 Standard ANSI/ANS Not Applicable Radiation Sources10 for CFR 50.49 (EQ) RG 1.143Conforms Partially Conforms in TableRG 1.112 See 1.9-2. 12.2.1 and RG 1.117 RG 1.29 RG 1.26, None.RG 1.7 ConformsConforms Partially None. in TableRG 1.143 See 1.9-2. 12.2.1 Not Applicable 11.2 11.1 3.2 II.B.2 AC II.7 II.3II.4 Operational ConsiderationsII.1 Radiation Protection Considerations Not Applicable Not II.3 II.2 II.4II.6 Task Action NUREG-0737, Item Plan II.1 II.8 II.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 12.2, Rev 0: Radiation DSRS Sources SRP 12.1, Rev 4: SRP 12.1, Rev Assuring That 4: Occupational Radiation As Low Is Exposures Are Reasonably Achievable SRP 12.1, Rev Assuring That 4: Occupational Radiation As Low Is Exposures Are Reasonably Achievable DSRS 12.2, Rev 0: Radiation DSRS Sources DSRS 12.2, Rev 0: Radiation DSRS Sources DSRS 12.2, Rev 0: Radiation DSRS Sources 12.2, Rev 0: Radiation DSRS Sources 12.2, Rev 0: Radiation DSRS Sources Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features DSRS 12.2, Rev 0: Radiation DSRS Sources DSRS 12.2, Rev 0: Radiation DSRS Sources
Tier 2 1.9-148 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 12.3.4 Comments Section See RG 1.52 in TableRG 1.52 See 1.9-2. in TableRG 1.69 See 1.9-2. in TableRG 1.97 See 1.9-2. Not Applicable in TableRG 1.183 See 1.9-2. 12.3.2 in TableRG 8.2 See 1.9-2. 7.2.13 in TableRG 8.8 See 1.9-2. 12.2 in TableRG 8.10 See 1.9-2. Not Applicable in TableRG 8.15 See 1.9-2. 12.3.1 Not Applicable in TableRG 8.25 See 1.9-2. Not Applicable in TableRG 8.38 See 1.9-2. Not Applicable 12.3.1 Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance RG 1.52RG 1.69RG 1.97RG 1.183 Applicable Not RG 8.2Conforms Partially RG 8.8Conforms Partially RG 8.10Conforms Partially RG 8.15 Not Applicable RG 8.19Conforms Partially RG 8.25 Applicable Not RG 8.38 Applicable Not ANSI/ANS/HPSSC-6.8.1-1981 Conforms None. Conforms Applicable Not None.Conforms Partially 12.4 12.3.4 AC II.2 II.3 II.4 II.5 II.6 II.7 II.8 II.9 II.10 II.12 II.13 II.11 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 12.3-12.4, Rev 0: Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features DSRS 12.3-12.4, Rev 0: Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features DSRS 12.3-12.4, Rev 0: Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features DSRS 12.3-12.4, Rev 0: Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features
Tier 2 1.9-149 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.4 12.3.4 emental extended LLW Comments Section The portion of this guidance that pertains to DCA. to the is applicable the design phase in TableRG 1.89 See 1.9-2. in TableRG 4.21 See 1.9-2. in TableRG 1.45 See 1.9-2. 3.11 12.3.6 guidance documents are not These applicableso far as they address the DCA to 5.2.5 the addition of suppl storage and the development of a PCP. This applicantis a COL responsibility. in TableRG 1.97 See 1.9-2. 12.3.4 Status Review Plan (SRP) and Design Specific Review Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance ANSI/HPS N13.1-2011ANSI/HPS ANSI/ANS-6.4-2006 ConformsRG 1.140 None. ConformsRG 1.89 None.RG 4.21Conforms Partially in TableRG 1.140 See 1.9-2.NEI 97-06Conforms Partially RG 1.143 12.3.4 Conforms Partially SECY-94-198 and BTP 11-3 12.3.3 12.3.2 Conforms Partially RG 1.97 ConformsConforms Partially None. in TableRG 1.143 See 1.9-2.Conforms Partially Ch 11 Ch 5 B. Matthews and Elmo E. Collins dated dated Collins E. Elmo and Matthews B. 10-10-2006 AC II.14 II.17 II.19 II.20II.21 RG 1.45II.24 Conforms Partially II.15 II.16 Memo from Larry W. Camper toDavid II.18 II.22 II.23 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 12.3-12.4, Rev 0: Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features DSRS 12.3-12.4, Rev 0: Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features DSRS 12.3-12.4, Radiation Protection Design Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features DSRS 12.3-12.4, Rev 0: Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features DSRS 12.3-12.4, Rev 0: Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features
Tier 2 1.9-150 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Comments Section oposed changes to existing emergency response facilities. See RG 1.12 in TableRG 1.12 See 1.9-2.This guidance governs operational programs, procedures, facilities and organization that are site-specific, and are the responsibilityCOL of the applicant referencing the certified design. COL applicant. 12.3.1 COL applicant.COL applicant.COL applicant.COL applicant. Not Applicable Not Applicable COL applicant. Not Applicable COL applicant.COL applicant. Not Applicable Not Applicable COL applicant.COL applicant. Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Status Review Plan (SRP) and Design Specific Review Not Applicable Not Applicable Not Applicable Not Applicablepr There are no Standard (DSRS) (Continued) AC Title/Description Conformance RG 1.12Conforms Partially Meeting the Standards of 10 CFR Full of Conduct 50.47(b); ParticipationExercise per 10 CFR 50, E Appendix Response Plans Level Scheme Facilities AC II.25 AllAll General and Specific RequirementsAll Not Applicable Operating OrganizationAll General and Specific RequirementsII.1 Not Applicable Not Applicable VariousII.2II.3II.4 Onsite and Offsite Emergency II.5 Emergency Classification and Action II.6 Meteorological Criteria Not Applicable II.7 Response Emergency Upgrading AlertingNotifications and Not Applicable Protective Action Recommendations Not Applicable Not Applicable All Various Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 12.3-12.4, Rev 0: Rev 0: 12.3-12.4, DSRS Design Radiation Protection Features 13.1.1, Rev 5: SRP Technical Management and Support Organization Rev 6: 13.1.3, 13.1.2 - SRP Organization Operating 13.2.1, Rev 3: Reactor SRP Operator Requalification Program; Reactor Operator Training 13.2.2, Rev 3: Non- SRP Training Staff Plant Licensed 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning DSRS 12.5, Rev 0: Operational Radiation Protection Program
Tier 2 1.9-151 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 13.3 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Comments Section applicants. the NuScale DCA. responseaspects data system. Site-specific are the responsibilityCOL of the applicant that references the NuScale certified design. applicants. applicants. applicants. applicants. applicants. applicants. COL applicant.COL applicant.COL applicant.COL applicant. Not Applicable COL applicant. Not Applicable COL applicant. Not Applicable Not Applicable Not Applicable Not Applicable orms The NuScale design includes an emergency Status Review Plan (SRP) and Design Specific Review Not Applicable Not Applicablesite Guidance applies permit to early Not Applicablepart of ITAAC are not Emergency planning Not Applicable Not Applicable Not Applicablesite Guidance applies permit to early Not Applicablesite Guidance applies permit to early Not Applicablesite Guidance applies permit to early Not Applicablesite Guidance applies permit to early Not Applicablesite Guidance applies permit to early Not Applicablesite Guidance applies permit to early with Early Site Early Site with Standard (DSRS) (Continued) AC Title/Description Conformance Alternatives to NUREG-0654/FEMA- 1, Rev REP-1, ITAAC Associatedwith Early Site Permit Application ITAAC Associatedwith Design Certification Application Permit Application Planning and Preparedness Local Governments Decline to Participate Characteristics Unique to Proposed Site Times of Evacuation Analysis Proposed Site Other Certifications Associated Plans and Plans AssociatedEarly Site with Permit Application AC II.8 II.9II.10II.11Government Tribal, and Local State, Planning Zones Emergency II.12 Estimates Evacuation Time Response Data System Emergency II.13 Not Applicable PartiallyConf II.14 Applicable Not Acceptability of Emergency PlansII.15 When Planning Emergency Offsite Not Applicable II.16Criteria - Physical Permit Early Site II.17Criteria - Preliminary Permit Early Site II.18 Physical Unique to Characteristics II.19or Agreement Copies of Letters II.20 Emergency Preparedness Information II.21 Emergency and Integrated Complete II.22 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 13.3, Rev 3: Emergency 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning
Tier 2 1.9-152 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 13.3 13.3 9.3.2 9.3.2 Not Applicable Not Applicable r displaysr are provided in Comments Section aspects of post-accident aspects of design includes a technical ic aspects of containmentic aspects of the technical support center.The emergencythe operationsis facility responsibility of the COL applicant that certification.referencesNuScale design the Content was subsumedSection into SRP 14.2. COL applicant.COL applicant. Not Applicable Not Applicable COL applicant.COL applicant.COL applicant.for this specific requirements There are no SRP section. COL applicant. Not Applicable Not Applicable Not Applicable Not Applicable support center. Thesupport center. operational support center and the emergency operations facility are the responsibilitythe COL of applicant that references the NuScale certified design. arethe responsibilitysamplingthe COL of applicant. arethe responsibilitysampling of the COL applicant. Status Review Plan (SRP) and Design Specific Review Not Applicable Not Applicable Not Applicable Not Applicable Partially Conforms The NuScale Partially Conforms Programmatic Partially Conforms Programmat System Conforms Safety paramete Standard (DSRS) (Continued) AC Title/Description Conformance ITAAC Associatedwith Combined License Application Generic Emergency Planning ITAAC Not Applicable and Notifications NRC Communications Various (Including Attachment, Sample FSAR Table 13.4-x) Emergency Response Facilities Containment Sampling from Potential Continuous Sampling Accident Release Points Pertaining to Orders Commission Emergency Planning AC II.23 II.24 II.25 Design and Implementation of II.26 Parameter Display Safety II.27II.28 Reactor Coolant System and II.29 Containmentand Monitoring II.30II.31 Generic Communications and Not Applicable Operational ProgramsAllAll Various Not Applicable Various Not Applicable Not Applicablefinalized.never Draft SRP section was Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 13.3, Rev 3: Emergency 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.3, Rev 3: Emergency SRP Planning 13.4, Rev 3: Operational SRP Programs SRP 13.5.1.1, Rev 1: Administrative Procedures - General SRP 13.5.1.2, Draft Rev 0: Administrative Procedures - Initial Test Program
Tier 2 1.9-153 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 14.2 Not Applicable Technical Report) 13.6.2 (via Security Comments Section nalized guidance in this SRP. Instead, applicablerelocated guidance was to SRP 17.5. COL applicant.COL applicant.Applicable for the physical security design certifiedelements within the boundary of the NuScale plant. Not Applicable ESP applicant.COL applicant.COL applicant. Not Applicable COL applicant.COL applicant. Not Applicable Not Applicable Not Applicable Not Applicable Applicant,Subheading Items A through DC D, are applicable to the DCA. Subheading Not Applicable are H, A through COL/OL applicants, Items applicableonly to COL applicant. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 14.2 14.2 Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance 10 CFR 73.56 Not Applicable Objectives Guides Regulatory Procedures AC II (no number) II (no AllAll VariousAll VariousAll VariousAll Various Not Applicable All VariousAll Not Applicable NRC never fi All Various Not Applicable II.1 Various Various Conforms II.2 Program and of Test Summary Not Applicable II.3 with Conformance Programs Test Not Applicable Not Applicable Test Program Administrative Initial Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 13.5.2.1, SRP 13.5.2.1, Rev 2: Operating and Emergency Operating Procedures SRP 13.5.2.2, Draft Rev 0: Maintenance and Other Operating Procedures 13.6.1, Rev 1: Physical SRP License Security - Combined and Operating Reactors 13.6.2, Rev 2: Physical SRP Certification Security - Design 13.6.3, Rev 1: Physical SRP Security - Early Site Permit 13.6.4, Rev 1: Access SRP Authorization 13.6.6, Rev 0: Cyber SRP Plan Security 13.7.1, Rev 0: Fitness for SRP Duty (Operational) 13.7.2, Rev 0: Fitness for SRP Duty (Construction) Plant 14.2, Rev 0: Initial DSRS - Design Test Program New COL Certification and applicants Plant 14.2, Rev 0: Initial DSRS - Design Test Program New COL Certification and applicants Plant 14.2, Rev 0: Initial DSRS - Design Test Program New COL Certification and applicants
Tier 2 1.9-154 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 14.2 14.2 14.3 Not Applicable Not Applicable Not Applicable Not Applicable Comments Section ology for developing ITAAC is developingology for ITAAC is developingology for provided in SRP 14.3. provided in SRP 14.3. Subheading DC Applicant,Subheading Item A, is DC COL/OL Subheading to the DCA. applicable to applicable B, are A and Items applicants, applicants. COL Applicant,Subheading Items A through DC applicableareSubheading C, to the DCA. COL/OL applicants,are Items A through C, applicableto COL applicants. This SRP section is applicable only to extended power uprate license amendment requests. criterionis of this acceptanceA portion applicableto COL applicants. ITAAC is developingMethodology for provided in SRP 14.3. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 14.2 14.3 Partially Conforms of ITAACof Conforms Partially Standard (DSRS) (Continued) AC Title/Description Conformance Individual Test Test Individual Descriptions/ Abstracts Acceptability of the Scope Specific Acceptance Criteria for ITAAC 14.3 Specified in SRP Section Criteria AC All Various Not Applicable II.4II.5 Initial Startup TestsII.6 Partially Conforms AllTest Program Acceptance Initial II.1 VariousII.2 All Not Applicable VariousAll Various Not Applicable Method Not Applicable Method Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 14.3.2, Rev 0: Structural 14.3.2, Rev 0: Structural SRP Engineering - and Systems Inspections, Tests, Analyses, and Acceptance Criteria DSRS 14.2, Rev 0: Initial Plant Plant 14.2, Rev 0: Initial DSRS - Design Test Program New COL Certification and applicants Plant 14.2, Rev 0: Initial DSRS - Design Test Program New COL Certification and applicants Plant 14.2, Rev 0: Initial DSRS - Design Test Program New COL Certification and applicants 14.2.1, (August 2006): SRP Generic Guidelines for Extended Power Uprate Testing Programs 14.3, (March 2007): SRP Inspections, Tests, Analyses, and Acceptance Criteria 14.3, (March 2007): SRP Inspections, Tests, Analyses, and Acceptance Criteria 14.3.3, (March 2007): SRP and Systems Piping Components - Inspections, Tests,and Analyses, Acceptance Criteria SRP 14.3.4, Rev 0: Reactor Systems - Inspections,Tests, Analyses,Acceptance and Criteria
Tier 2 1.9-155 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 14.3.12 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Comments Section ology for developing ITAAC is developingology for ITAAC is developingology for L applicantPhysical addresses provided in SRP 14.3. provided in SRP 14.3. Methodology for developing ITAAC is developingMethodology for provided in SRP 14.3. ITAAC is developingMethodology for provided in SRP 14.3. ITAAC is developingMethodology for provided in SRP 14.3. ITAAC is developingMethodology for provided in SRP 14.3. ITAAC is developingMethodology for provided in SRP 14.3. Security Hardware ITAAC outside of the of the outside Hardware ITAAC Security nuclear island and structures. Status Review Plan (SRP) and Design Specific Review Conforms None. 15.0 Standard (DSRS) (Continued) AC Title/Description Conformance Accidents AC All Various Not Applicable AllAll VariousAll VariousAll Various Not Applicable Various Not Applicable Method Not Applicable Method Partially Conforms The CO AllAll VariousAll Various Various Not Applicable Not Applicable Not Applicable I.1 Categorization of Transients and Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 14.3.5, Rev 0: Instrumentation and Controls - Inspections, Tests,Analyses, and Acceptance Criteria 14.3.9, (March 2007): SRP - Human Factors Engineering Inspections, Tests, Analyses, and Acceptance Criteria SRP 14.3.10, (March 2007): Emergency Planning - Inspections, Tests, Analyses, and Acceptance Criteria SRP 14.3.11, (March 2007): Containment Systems - Inspections, Tests, Analyses, and Acceptance Criteria SRP 14.3.12, Rev 1: Physical Security Hardware - Inspections, Tests, Analyses, and Acceptance Criteria SRP 14.3.6, Rev 0: Electrical 14.3.6, Rev 0: Electrical SRP Systems - Inspections,Tests, Analyses,Acceptance and Criteria 14.3.7, Rev 0: Plant SRP Systems - Inspections,Tests, Analyses,Acceptance and Criteria 14.3.8, Rev 0: Radiation SRP Protection - Inspections, Tests,and Analyses, Acceptance Criteria DSRS 15.0, Rev 0: Introduction Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses
Tier 2 1.9-156 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.0 15.0 15.0.0 Comments Section applicable except for 4.B.ii ve been historically classified and 4.B.iv. CHF, not DNBR, is used to not DNBR, is used 4.B.iv. CHF, and fuel margin for the thermal determine the acceptancecladding. criteria uses LOCA an is more restrictiveacceptancecriterion that 2,200 degree than the temperature limit of F. This acceptance criterion is applicable This applicableexcept for Item B.vi, which is to applicants. COL as AOOs are not analyzedfrequency for of occurrence. Some events that have an IE are also deterministically frequency AOOs. as classified Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None.Conforms None. 15.0 15.0 15.0 15.0 Partially ConformsPartially The guidance is Partially Conforms Partially Conforms Events that ha for AOOs Conforms None. 15.0.0 Standard (DSRS) (Continued) AC Title/Description Conformance Analysis Acceptance Criteria Analysis Acceptancefor IEs Criteria and Postulated Accidents the Safety Evaluation the Safety Systems Actions Events Frequency Classification Operation Frequency of Occurrence AC I.5I.6 PlantConsidered in Characteristics I.7 Assumed Protection and Safety I.8.A of Individual Initiating Evaluation I.8.Bof Causes and Identification Sequenceof Events and Systems I.2 Categorization According to I.4.B I.3I.4.A Categorization According to Type Conforms None. 15.0 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.0, Rev 0: Introduction Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses DSRS 15.0, Rev 0: Introduction Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses DSRS 15.0, Rev 0: Introduction Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses DSRS 15.0, Rev 0: Introduction Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses
Tier 2 1.9-157 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.0.2 15.0.2 15.0.2 15.0.2 Not Applicable Not Applicable sensitivity analyses flux (CHF), which is Comments Section The guidance is applicableThe guidance is except for aspectsNuScalethat are BWR-specific. evaluates critical heat more applicable to the NuScale design than DNBR. The NuScale design uses a modified version (AST) the alternative source term of methodologyradiological to evaluate consequences of accidents. The NuScale design utilizes a modified version of the AST methodology to evaluate radiological consequences of accidents. design supports an exemptionThe NuScale 10selected portions of from CFR 50 Appendix evaluation K. NuScale ECCS models for LOCAs only address technically by Appendix K. relevant features required design supports an exemptionThe NuScale 10selected portions of from CFR 50 Appendix evaluation K. NuScale ECCS models for LOCAs only assessthe technically relevant features required by II.K3.30. Item Appendix K and TMI Action Non-LOCA methods use to determine input or bounding values parameters. Status Review Plan (SRP) and Design Specific Review Conforms None. 15.0.2 Not Applicable Not Applicable Partially Conforms Standard (DSRS) (Continued) h on Page 15.0.1-6 and 15.0.1-6 h on Page AC Title/Description Conformance Performance Compliance with Page 15.0.1-6, Provisions of NUREG-0737 Specific Process Table 1, Criteria Exposure for Radiological Consequences of Design Basis Accident AC I.8.C Core, System, and Barrier II (No number) First fulland paragraph 6 bullets on II (No number) Last paragrap II.1 Evaluation ModelII.2II.3 Accident Scenario Identification Code Assessment Departure II.4II.5 Uncertainty Analysis Departure Quality Assurance Plan Conforms Conforms None. 15.0.2 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.0, Rev 0: Introduction Introduction 0: Rev 15.0, DSRS Accident - Transient and Analyses Rev 0: Radiological SRP 15.0.1, Using Consequence Analyses Terms Alternative Source Rev 0: Radiological SRP 15.0.1, Using Consequence Analyses Terms Alternative Source 15.0.2, (March 2007): SRP of Transient and Review Accident Analysis Methods 15.0.2, (March 2007): SRP of Transient and Review Accident Analysis Methods 15.0.2, (March 2007): SRP of Transient and Review Accident Analysis Methods 15.0.2, (March 2007): SRP of Transient and Review Accident Analysis Methods 15.0.2, (March 2007): SRP of Transient and Review Accident Analysis Methods
Tier 2 1.9-158 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.4 13.3 15.0.3 15.0.3 Comments Section ance criterion met for TSC when AC power is available. TSC function is TSC power is available. when AC transferred to the main control room when AC is not available. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 15.0.3 6.4 Conforms None.Conforms None. 15.1.1-15.1.4 15.1.1-15.1.4 Partially Conforms Dose accept Standard (DSRS) (Continued) AC Title/Description Conformance Postulated DBAs (includes Table 1) Table DBAs (includes Postulated Habitability Radiological Habitability Identify Limiting Increase in Heat Removal Events and System Verify Fuel Damage for Limiting Pressure Criteria are Met Event. AC II.1II.2 Offsite Radiological Consequences of Radiological Room Control II.3II.1 Technical Support Center Basic Objective II.2 Basic Objective Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.0.3, Rev 0: Design Accident Radiological Basis for Consequence Analyses SMR Design the NuScale DSRS 15.0.3, Rev 0: Design Accident Radiological Basis for Consequence Analyses SMR Design the NuScale DSRS 15.0.3, Rev 0: Design Accident Radiological Basis for Consequence Analyses SMR Design the NuScale DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System
Tier 2 1.9-159 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.1.1-15.1.4 15.1.1-15.1.4 ly describes plant plant describes ly Comments Section NuScale has determined that criticalNuScale has determined that heat flux more accurate phenomena than departure from nucleate boiling. NuScale events are classified by AOO, IE, accident, andbut special will event, requirementconform with SRP that incidents of moderate frequency should not generatea more seriousplant condition without other faults occurring independently. Status Review Plan (SRP) and Design Specific Review Conforms Conforms None. 15.1.1-15.1.4 Standard (DSRS) (Continued) AC Title/Description Conformance MCHFR Remains Above 95/95 Limit MCHFR Remains Above 95/95 Conforms Generate More Not AOOs Should Serious Condition Setpoints use Instrument Spans and RG 1.105 System Pressure Conforms None. 15.1.1-15.1.4 AC II.1 Specific Criterion II.3 Specific Criterion II.4 Specific Criterion II.2 Specific Criterion Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.1.1-15.1.4, DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System DSRS 15.1.1-15.1.4, DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System DSRS 15.1.1-15.1.4, DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System
Tier 2 1.9-160 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance Identify Limiting Single Failure Conforms None. 15.1.1-15.1.4 Initial Power Level is 102% Conforms None. 15.1.1-15.1.4 Conservative Scram Used Conforms None. 15.1.1-15.1.4 Core Burnup Conforms None. 15.1.1-15.1.4 AC II.5 Specific Criterion II.1 Analytical II.1 Analytical Parameters II.2 Analytical II.2 Analytical Parameters II.3 Analytical II.3 Analytical Parameters Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.1.1-15.1.4, DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System DSRS 15.1.1-15.1.4, DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System DSRS 15.1.1-15.1.4, DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System DSRS 15.1.1-15.1.4, DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System
Tier 2 1.9-161 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.1.5 ly describes plant plant describes ly Comments Section NuScale has determined that criticalNuScale has determined that heat flux more accurate phenomena than departure from nucleate boiling. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 15.1.1-15.1.4 15.1.5 Conforms None.Conforms None.Conforms None. 15.1.5 15.1.5 15.1.5 Conforms None. 15.1.5 Standard (DSRS) (Continued) AC Title/Description Conformance Reactor Coolant andReactor Coolant Main Steam System Pressure of Core Damage PotentialEvaluation Conforms Setpoint Inaccuracies use guidance in RG 1.105 RadiologicalLine for Steam Criteria Breaks Safety-Related Classification and Heat of Decay Auto-Initiation Removal System Initial Power Level and Plant Operating Mode PowerLoss of Offsite Conforms None. 15.1.5 Postulated Steam Line Break Effects Steam Line Postulated Conforms None. 15.1.5 Worst Case Failure of Single Active Component AC II.1 Specific Criteria II.2 Specific Criteria II.4 Analytical II.4 Analytical Parameters II.3 Specific Criteria II.4 Specific Criteria II.1 Assumptions II.2 Assumptions II.3 Assumptions II.4 Assumptions Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.1-15.1.4, DSRS 15.1.1-15.1.4, Rev 0: Decrease in Feedwater Temperature, Increase in Increase in Feedwater Flow, and Inadvertent Steam Flow, Opening of the Turbine Bypass System orInadvertent Operation of the Decay Heat Removal System DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment
Tier 2 1.9-162 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.0.3 15.1.5 Not Applicable ns, SRP Section 15.0.3 ns, SRP ly describes plant plant describes ly Comments Section s determined that criticals determined that heat Operator Action is not required to mitigate Operator Action is not ofthe consequences a steam linebreak. of Areas Section I, Section 15.0.3, SRP Per Review, under subheading Item 10 Review for the review of designInterfaces, certification applicatio radiologicalsupersedes the analyses, assumptions,acceptance criteria, and methodologies identified in this SRP Provisions 15.1.5, Appendix A. Section related to thenonradiological analyses aspects of this SRP Section 15.1.5, Appendix applicableA, remain to the DCA. flux more accurate phenomena than departure from nucleate boiling. Status Review Plan (SRP) and Design Specific Review Conforms None. 15.1.5 Conforms None.Conforms None. 15.1.5 15.1.5 Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Maximum-Worth Rod Fully Withdrawn Core Burnup Conforms None. 15.1.5 Initial Core Flow Conforms NuScale ha Postulated Failure of Non-Seismic Line Main Steam Postulated Failure of Seismic Main Steam Line Limiting Consequence Assessment Action is Credited When Operator AC II.5 Assumptions II.6 Assumptions All Various Partially Conforms II.7 Assumptions II.8 Assumptions II.9 Assumptions II.10 Assumptions Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment Rev 2: 15.1.5.A, SRP Radiological Consequences of Main Steam Line Failures ContainmentOutside of a PWR DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment DSRS 15.1.5, Rev 0: Steam System PipingInside Failures and Outside of Containment
Tier 2 1.9-163 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.0.3 15.0.3 Not Applicable Comments Section The part of this guidance specifying the guidanceThe part specifying of this of consequences radiological of calculation outside a postulated main steam linebreak containment is applicable to the DCA. I, 15.0.3, Section However, per SRP Section Areas of Review, Item 10.A under Review Interfaces,part thesubheading of specifies this acceptance criterion that acceptance criteria is radiological Section 15.0.3. superseded by SRP acceptance criterion specifiesThis radiological analysis methodology and SRP that are supersededassumptions by Section 15.0.3. The partguidance of this related to the specification for primary required technical secondary coolantand iodine activityand primary-to-secondary leak rate is applicable Section However, per SRP the DCA. to Section15.0.3, I, Areas of Review, Item 10.A Review Interfaces, theunder subheading of this acceptancepart criterion that radiological acceptancespecifies is criteria Section 15.0.3. superseded by SRP Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 15.1.6 15.1.6 Not Applicable Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance on Page 15.1.5-11, Exposure 15.1.5-11, on Page Guidelines for Calculated Doses 15.1.5-11, and 2 on Page Methodology and Assumptions for Calculating Radiological Consequences 15.1.5-11, and 2 on Page Items 1 for Assumed Technical Specifications Primary-to- and Iodine Activity Leak Rate Secondary Reactor Coolant PressureReactor Coolant Conforms None. 15.1.6 Cladding Integrity Conforms None. 15.1.6 AOOs Should Not Generate More Generate More Not AOOs Should Serious Condition use Setpoints and Spans Instruments RG 1.105 Identify Limiting Single Failure Conforms None. 15.1.6 AC II (No number) and 2 1 and Items paragraph full First II (No number)Items 1 following paragraph full First II (No number)following paragraph Second full II.1 Specific II.1 Specific Criteria II.2 Specific II.2 Specific Criteria II.3 Specific II.3 Specific Criteria II.4 Specific Criteria II.5 Specific Criteria Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.1.5.A, Rev 2: Rev 2: 15.1.5.A, SRP Radiological Consequences of Main Steam Line Failures ContainmentOutside of a PWR Rev 2: 15.1.5.A, SRP Radiological Consequences of Main Steam Line Failures ContainmentOutside of a PWR Rev 2: 15.1.5.A, SRP Radiological Consequences of Main Steam Line Failures ContainmentOutside of a PWR DSRS 15.1.6, Rev 0: Loss of Containment Vacuum DSRS 15.1.6, Rev 0: Loss of Containment Vacuum DSRS 15.1.6, Rev 0: Loss of Containment Vacuum DSRS 15.1.6, Rev 0: Loss of Containment Vacuum DSRS 15.1.6, Rev 0: Loss of Containment Vacuum
Tier 2 1.9-164 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.2.1-15.2.5 Comments Section The NuScale design does not have a steam pressure regulator. Status Review Plan (SRP) and Design Specific Review Conforms None. 15.2.1-15.2.5 Conforms None.Conforms None. 15.1.6 15.1.6 Partially Conforms Events Conforms None. 15.2.1-15.2.5 Standard (DSRS) (Continued) AC Title/Description Conformance Initial Power Level is 102% Conforms None. 15.1.6 Steam System Pressures Conservative Scram Used Conforms None. 15.1.6 Core Burnup Conforms None. 15.1.6 Maximize Heat Transfer from RCS to from RCS Heat TransferMaximize Containment and Reactor Pool Setpoint InaccuraciesGuidancein use RG 1.105 Moderate Frequency AC II.1 Analytical II.1 Analytical Parameters II.2.A Reactor Coolant System and Main II.2 Analytical II.2 Analytical Parameters II.3 Analytical II.3 Analytical Parameters II.4 Analytical II.4 Analytical Parameters II.5 Analytical Parameters II.1 Basic Objectives - Initiating II.2 SpecificCriteria for Events of Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Load; Turbine Load; Turbine Load; Turbine Title SRP or DSRS Section, Rev: DSRS 15.1.6, Rev 0: Loss of Containment Vacuum DSRS 15.2.1-15.2.5, DSRS 15.2.1-15.2.5, Rev 0: Loss of External Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.1.6, Rev 0: Loss of Containment Vacuum DSRS 15.1.6, Rev 0: Loss of Containment Vacuum DSRS 15.1.6, Rev 0: Loss of Containment Vacuum DSRS 15.1.6, Rev 0: Loss of Containment Vacuum DSRS 15.2.1-15.2.5, Rev 0: Loss of External Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.2.1-15.2.5, DSRS 15.2.1-15.2.5, Rev 0: Loss of External Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed)
Tier 2 1.9-165 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.2.1-15.2.5 Comments Section The NuScale design does not have a steam pressure regulator. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 15.2.1-15.2.5 15.2.1-15.2.5 15.2.1-15.2.5 quency Conforms None. 15.2.1-15.2.5 Standard (DSRS) (Continued) AC Title/Description Conformance Plant Response Failure Failures of Active Single Systems and Systems and Passive AC II.2.B Integrity Cladding Fuel Conforms II.2.CFre Incidents of Moderate II.2.Don Setpoints - Impact Instrument II.2.E Most Limiting PlantSystem Single II.2.F Performance of Nonsafety-Related Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Load; Turbine Load; Turbine Load; Turbine Load; Turbine Load; Turbine Title SRP or DSRS Section, Rev: Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.2.1-15.2.5, DSRS 15.2.1-15.2.5, Rev 0: Loss of External DSRS 15.2.1-15.2.5, DSRS 15.2.1-15.2.5, Rev 0: Loss of External Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.2.1-15.2.5, Rev 0: Loss of External Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.2.1-15.2.5, Rev 0: Loss of External Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.2.1-15.2.5, DSRS 15.2.1-15.2.5, Rev 0: Loss of External
Tier 2 1.9-166 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Comments Section Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 15.2.1-15.2.5 Conforms None. 15.2.1-15.2.5 15.2.1-15.2.5 15.2.1-15.2.5 Standard (DSRS) (Continued) AC Title/Description Conformance Analytical Model - Initial Power Level and Modesof Operation Analytical Model - Scram Characteristics Analytical Model - Core Burnup Analytical Model - Instrumentation System for Mitigating Setpoints Actuation AC II.3 Analytical ModelII.3.A Values of Parameters Used in ConformsII.3.B None. Values of Parameters Used in II.3.C Values of Parameters Used in II.3.D 15.2.1-15.2.5 Values of Parameters Used in Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Load; Turbine Load; Turbine Load; Turbine Load; Turbine Load; Turbine Title SRP or DSRS Section, Rev: Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.2.1-15.2.5, Rev 0: Loss of External Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.2.1-15.2.5, Rev 0: Loss of External Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.2.1-15.2.5, Rev 0: Loss of External Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.2.1-15.2.5, Rev 0: Loss of External Trip;of Loss Condenser Vacuum; Closure of Main Steam Isolation Valve; and Steam Pressure Regulator Failure (Closed) DSRS 15.2.1-15.2.5, DSRS 15.2.1-15.2.5, Rev 0: Loss of External
Tier 2 1.9-167 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.2.8 Comments Section uses a heat generationrate limit to ensure uses a heat that fuel centerline melting limits are met. NuScale categorizes events as AOO and IE. 15.2.6 NuScale categorizes events as AOO and IE. 15.2.7 rms For slower reactivity insertions, NuScale Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 15.2.6 15.2.7 Conforms None.Conforms None. 15.2.6 Conforms None. 15.2.6 15.2.8 thods Conforms None. 15.2.7 Standard (DSRS) (Continued) AC Title/Description Conformance met Requirements of GDC 10 and GDC 15 and GDC 10 Requirements of GDC Conforms None. Power - of AC of Loss Analysis Model Analytical and Methods, conservative assumptions and RG 1.105 15.2.6 System Pressures Failure Steam System Pressures AC II.1II.2 Fuel and System Pressure Parameters Events of Moderate Frequency Conforms II.1II.2 Reactorand Coolant Main Steam II.3 Integrity Cladding Fuel II.4 Frequency Incidents of Moderate II.5 Conforms ConformsII.5 A-D None. Most Limiting PlantSystem Single II.1II.2 Reactor Coolant System and Main II.3 15.2.6 Evaluation of Core Damage Potential Confo Doses Calculated Site Boundary Conforms None. 15.2.8 II.3 Model and Me Analytical Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.2.7, Rev 0: Loss of Normal Feedwater Flow DSRS 15.2.7, Rev 0: Loss of Normal Feedwater Flow DSRS 15.2.6, Rev 0: Loss of Power to Nonemergency AC Auxiliaries Station the DSRS 15.2.6, Rev 0: Loss of Power to Nonemergency AC Auxiliaries Station the DSRS 15.2.6, Rev 0: Loss of Power to Nonemergency AC Auxiliaries Station the DSRS 15.2.6, Rev 0: Loss of Power to Nonemergency AC Auxiliaries Station the DSRS 15.2.6, Rev 0: Loss of Power to Nonemergency AC Auxiliaries Station the DSRS 15.2.6, Rev 0: Loss of Power to Nonemergency AC Auxiliaries Station the 15.2.8, Rev 0: Feedwater DSRS System PipeBreakand Inside Outside Containment (PWR) 15.2.8, Rev 0: Feedwater DSRS System PipeBreakand Inside Outside Containment (PWR) 15.2.8, Rev 0: Feedwater DSRS System PipeBreakand Inside Outside Containment (PWR) DSRS 15.2.7, Rev 0: Loss of Normal Feedwater Flow
Tier 2 1.9-168 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.4.1 15.4.1 15.4.2 Not Applicable Not Applicable Not Applicable eapplicable only to ow for core cooling. Comments Section SRP states that this criterion applies to Applicablerelydesigns onlythat on to LWR forced reactor coolant fl The NuScale design uses passive natural circulation of the primary coolant, eliminating the need for reactor coolant pumps. Section 15.3.3 - 15.3.4 ar LWR designs that rely on forced reactor flowcoolantcore for cooling. The NuScale design uses passive natural circulation of the primary coolant, eliminating the need for reactor coolant pumps. flux (CHF) is more appropriate Critical heat terminology for NuScale phenomena than departure from nucleate boiling (DNBR). For slower reactivity insertions, NuScale generationrate limit to ensure uses a heat that fuel centerline melting limits are met. for CHF is more appropriate terminology NuScale phenomenon than DNBR. BWRs. NuScale uses the 95/95 MCHFR BWRs. approachensure tocladding no or fuel failures. Status Review Plan (SRP) and Design Specific Review Conforms None. 15.2.8 Conforms None. 15.2.8 Standard (DSRS) (Continued) stulated Failures r Initial Plant AC Title/Description Conformance Conditions and Po DHRS must be safetygrade and automatically initiated when required. AC II.4 II.5All fo Assumptions VariousAll Various Not Applicable Not Applicable II.1.A Thermal Margin LimitsII.1.B Fuel Centerline TemperaturesII.1.C Conforms Conforms UniformStrain Cladding II.1.A Thermal Margin Limits Not Applicable The Conforms Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.2.8, Rev 0: Feedwater DSRS System PipeBreakand Inside Outside Containment (PWR) 15.2.8, Rev 0: Feedwater DSRS System PipeBreakand Inside Outside Containment (PWR) SRP 15.3.1-15.3.2, Rev 2: Loss Coolant of Forced Reactor Pump Flow Including Trip of Flow Controller Motor and Malfunctions SRP 15.3.3-15.3.4, Rev 2: Pump Rotor Reactor Coolant Reactor CoolantSeizure and Pump Shaft Break SRP 15.4.1, Rev 3: SRP Uncontrolled Control Rod Assembly a From Withdrawal Subcritical or Low Power Startup Condition 15.4.1, Rev 3: SRP Uncontrolled Control Rod Assembly a From Withdrawal Subcritical or Low Power Startup Condition 15.4.1, Rev 3: SRP Uncontrolled Control Rod Assembly a From Withdrawal Subcritical or Low Power Startup Condition 15.4.2, Rev 3: SRP Uncontrolled Control Rod Assemblyat Withdrawal Power
Tier 2 1.9-169 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.4.2 15.4.3 15.4.3 Not Applicable Not Applicable Comments Section SRP states that this criterion applies to guidance is not applicable because the r slower reactivity insertions, NuScale For slower reactivity insertions, NuScale generationrate limit to ensure uses a heat that fuel centerline melting limits are met. BWRs. NuScale uses the 95/95 MCHFR BWRs. approachensure tocladding no or fuel failures. that designs PWR to refers language specific flow and have forced reactor coolant use reactor coolant loopsand pumps. The NuScale design does not require or include reactor coolant pumps.The potential for a reactivity accidentpostulated startup (e.g., initiated by abnormal startup sequence) has requiringasbeen identified an event NuScaleconsideration for the reactor design. The guidance for these AOOs is available in DSRS 15.4.6 - Inadvertent DecreaseConcentration in Boron in the Reactor Coolant System (PWR). CHF is more appropriate terminology for CHF is more appropriate terminology NuScale phenomenon than DNBR. generationrate limit to meet uses a heat fuel centerlinemelting limits. Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance AC II.1.B Fuel Centerline TemperaturesII.1.C UniformStrain Cladding II.1 Conforms Thermal Margin Limits Not Applicable The Conforms II.A Pressures MSS RCS and Applicable Not This II.2II.3 Fuel Centerline Temperatures UniformStrain Cladding Conforms Fo Conforms None. 15.4.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.4.2, Rev 3: SRP Uncontrolled Control Rod Assemblyat Withdrawal Power SRP 15.4.4-15.4.5, Rev 2: orLoop Startup of an Inactive atRecirculation Loop an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate SRP 15.4.2, Rev 3: SRP Uncontrolled Control Rod Assemblyat Withdrawal Power Rev 3: SRP 15.4.3, Rod Control Misoperation (System Malfunction or Operator Error) SRP 15.4.3, Rev 3: SRP 15.4.3, Rod Control Misoperation (System Malfunction or Operator Error) Rev 3: SRP 15.4.3, Rod Control Misoperation (System Malfunction or Operator Error)
Tier 2 1.9-170 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Comments Section guidance is not applicable because the specific language refers to PWR designs that that designs PWR to refers language specific flow and have forced reactor coolant use reactor coolant loopsand pumps. The NuScale design does not require or include reactor coolant pumps.The potential for a reactivity accidentpostulated startup (e.g., initiated by abnormal startup sequence) has requiringasbeen identified an event NuScaleconsideration for the reactor design. The guidance for these AOOs is available in DSRS 15.4.6 - Inadvertent DecreaseConcentration in Boron in the Reactor Coolant System (PWR). This guidance is not applicable because the that designs PWR to refers language specific flow and have forced reactor coolant use reactor coolant loopsand pumps. The NuScale design does not require or include reactor coolant pumps.The potential for a reactivity accidentpostulated startup (e.g., initiated by abnormal startup sequence) has requiringasbeen identified an event NuScaleconsideration for the reactor design. The guidance for these AOOs is available in DSRS 15.4.6 - Inadvertent DecreaseConcentration in Boron in the Reactor Coolant System (PWR). Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance AC II.B Fuel Thermal Limits Not Applicable This II.C Events of Moderate Frequency Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.4.4-15.4.5, SRP 15.4.4-15.4.5, Rev 2: orLoop Startup of an Inactive atRecirculation Loop an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate SRP 15.4.4-15.4.5, SRP 15.4.4-15.4.5, Rev 2: orLoop Startup of an Inactive atRecirculation Loop an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate
Tier 2 1.9-171 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Comments Section idance is not applicable because the specific language refers to PWR designs that that designs PWR to refers language specific flow and have forced reactor coolant use reactor coolant loopsand pumps. The NuScale design does not require or include reactor coolant pumps.The potential for a reactivity accidentpostulated startup (e.g., initiated by abnormal startup sequence) has requiringasbeen identified an event NuScaleconsideration for the reactor design. The guidance for these AOOs is available in DSRS 15.4.6 - Inadvertent DecreaseConcentration in Boron in the Reactor Coolant System (PWR). specific language refers to PWR designs that that designs PWR to refers language specific flow and have forced reactor coolant use reactor coolant loopsand pumps. The NuScale design does not require or include reactor coolant pumps.The potential for a reactivity accidentpostulated startup (e.g., initiated by abnormal startup sequence) has requiringasbeen identified an event NuScaleconsideration for the reactor design. The guidance for these AOOs is available in DSRS 15.4.6 - Inadvertent DecreaseConcentration in Boron in the Reactor Coolant System (PWR). Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance AC II.D Instrument Setpoints Not Applicableapplicablenotthe guidance is This because II.E Single Failure Not Applicable This gu Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.4.4-15.4.5, SRP 15.4.4-15.4.5, Rev 2: orLoop Startup of an Inactive atRecirculation Loop an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate SRP 15.4.4-15.4.5, SRP 15.4.4-15.4.5, Rev 2: orLoop Startup of an Inactive atRecirculation Loop an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate
Tier 2 1.9-172 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.4.6 Not Applicable Not Applicable Comments Section guidance is not applicable because the an inadvertentdilution boron event. CHF is more appropriate terminology for CHF is more appropriate terminology NuScale phenomenon. specific language refers to PWR designs that that designs PWR to refers language specific flow and have forced reactor coolant use reactor coolant loopsand pumps. The NuScale design does not require or include reactor coolant pumps.The potential for a reactivity accidentpostulated startup (e.g., initiated by abnormal startup sequence) has requiringasbeen identified an event NuScaleconsideration for the reactor design. The guidance for these AOOs is available in DSRS 15.4.6 - Inadvertent DecreaseConcentration in Boron in the Reactor Coolant System (PWR). Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms 15.4.6 None. 15.4.6 Not Applicable Operator action is not required to mitigate Standard (DSRS) (Continued) rvals for Required rvals for AC Title/Description Conformance Operator Actions System Pressures Assumptions AC II.F Non-Safety Systems Not Applicable This II.1II.2 Reactorand Coolant Main Steam II.3 Integrity Cladding Fuel II.4Frequency Incidents of Moderate Conforms ConformsII.5 Minimum Time Inte None. Analysis Model, Methods, and 15.4.6 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.4.4-15.4.5, SRP 15.4.4-15.4.5, Rev 2: orLoop Startup of an Inactive atRecirculation Loop an Incorrect Temperature, and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate SRP 15.4.6, Rev 2: Inadvertent Decrease in Boron the Reactor in Concentration Coolant System (PWR) SRP 15.4.6, Rev 2: Inadvertent Decrease in Boron the Reactor in Concentration Coolant System (PWR) SRP 15.4.6, Rev 2: Inadvertent Decrease in Boron the Reactor in Concentration Coolant System (PWR) SRP 15.4.6, Rev 2: Inadvertent Decrease in Boron the Reactor in Concentration Coolant System (PWR) SRP 15.4.6, Rev 2: Inadvertent Decrease in Boron the Reactor in Concentration Coolant System (PWR)
Tier 2 1.9-173 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.4.7 15.0.3 15.0.3 ns, SRP Section 15.0.3 ns, SRP of this acceptance Comments Section Safety analysisdemonstratesSafety that there are failures. no fuel of Areas Section I, Section 15.0.3, SRP Per Review, under subheading Item 10 Review for the review of designInterfaces, certification applicatio radiologicalsupersedes the analyses, assumptions,acceptance criteria, and methodologies identified in SRP Section the related to Appendix A. Provisions 15.4.8, nonradiological analysesthis SRP aspects of to the apply 15.4.8, Appendix A, Section DCA. the guidanceThe part specifying of this of consequences radiological of calculation a postulated control rod ejection accident is per SRP However, to the DCA. applicable I, Areas of Review, 15.0.3, Section Section under subheading Item 10.C Review Interfaces, the part criterion that specifies radiological byacceptancecriteria is superseded SRP Section 15.0.3. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 15.4.7 15.4.8 15.4.8 Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance First paragraph of Section II (bottom Section of First paragraph of pagepage 15.4.8-5 and top of and of Site15.4.8-6) - Acceptability ESF Dose Mitigating Procedures Requiring to Detect Fuel Instrumentation Loading Errors Instrumentation Accidents AC II.1II.2 Provision in Plant Operating II.1 Offsite Radiological ConsequencesII.2 Conforms II.3Monitoring Availability of All Reactivity Postulated of Effects Radiation Dose Limits Various ConformsII (No number) None. Partially Conforms 15.4.8 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.4.7, Rev 2: Inadvertent Operation of a Loading and Fuel Assembly in an Improper Position SRP 15.4.7, Rev 2: Inadvertent Operation of a Loading and Fuel Assembly in an Improper Position SRP 15.4.8, Rev 3: Spectrum of Rod Ejection Accidents (PWR) SRP 15.4.8, Rev 3: Spectrum of Rod Ejection Accidents (PWR) SRP 15.4.8, Rev 3: Spectrum of Rod Ejection Accidents (PWR) Rev 1: 15.4.8.A, SRP Radiological Consequences Ejection Rod Control a of Accident (PWR) Rev 1: 15.4.8.A, SRP Radiological Consequences Ejection Rod Control a of Accident (PWR)
Tier 2 1.9-174 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.0.3 15.5.1-15.5.2 Not Applicable Not Applicable Not Applicable Comments Section The partguidance of this related to the specification for primary- required technical to the leak rate is applicable to-secondary However, per SRP Section 15.0.3, DCA. I, Areas of Review, under Section Item 10.C Review Interfaces,part thesubheading of specifies this acceptance criterion that acceptance criteria is radiological Section 15.0.3. superseded by SRP of Areas Section I, Section 15.0.3, SRP Per under subheading Review, Item 10.C Review Interfaces, this acceptance criterion radiological acceptancespecifies criteria assumptionsand that are supersededby 15.0.3. SRP Section criteriasection and its acceptance SRP This through are applicable (II.1 II.3) only to BWRs. criteriasection and its acceptance SRP This are applicableonly to BWRs. Status Review Plan (SRP) and Design Specific Review Conforms None. 15.0 Conforms None.Conforms None. 15.5.1-15.5.2 15.5.1-15.5.2 Not Applicable Partially Conforms Standard (DSRS) (Continued) g assumptions for for g assumptions ragraph on page 15.4.8-6) on page 15.4.8-6) ragraph AC Title/Description Conformance The frequency classificationfor this event is anAOO. - Technical Specification for Primary- for - Technical Specification to-Secondary Leak Rate Model 15.4.8-6) - Dose initiationuntil a stabilized condition includin is reached operates,equipment fails to that operate or requires operator action. approved model or be justified. AC II (No number) First full pa II (No number)on page paragraph Second full AllAll - - Applicable Not Applicable Not II.1 II.2 The sequence of events, from II.3 an be must Model Evaluation Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.4.8.A, Rev 1: Rev 1: 15.4.8.A, SRP Radiological Consequences Ejection Rod Control a of Accident (PWR) Rev 1: 15.4.8.A, SRP Radiological Consequences Ejection Rod Control a of Accident (PWR) SRP 15.4.9, Rev 3: Spectrum of (BWR) Accidents Rod Drop Draft Rev 3: 15.4.9.A, SRP Radiological Consequences Accident Drop Rod Control of (BWR) DSRS 15.5.1-15.5.2, DSRS 15.5.1-15.5.2, Rev 0: Chemical and Volume Control System Malfunction That IncreasesReactor Coolant Inventory DSRS 15.5.1-15.5.2, DSRS 15.5.1-15.5.2, Rev 0: Chemical and Volume Control System Malfunction That IncreasesReactor Coolant Inventory DSRS 15.5.1-15.5.2, Rev 0: Chemical and Volume Control System Malfunction That IncreasesReactor Coolant Inventory
Tier 2 1.9-175 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.6.1 15.6.6 actuation, analyzed Comments Section This guidance is only applicable to LWRs that are designed with power-operated pressurizer relief valves. The NuScale design relief valves use power-operated does not (PORVs), whichhave the potential to open inadvertently. Rather, the NuScale design relief code safety ASME springloaded uses notvalves, which do have the PORVs operation. vulnerability to inadvertent However, a mechanical failure of the reactor safety valve (RSV) is bounded by an inadvertent ECCS valve 15.6.6. in Section Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None.Conforms None. 15.5.1-15.5.2 15.5.1-15.5.2 15.5.1-15.5.2 Standard (DSRS) (Continued) AC Title/Description Conformance Conditions - Initial Power Level Conditions - Scram Characteristics Conditions - Core Burnup Conditions AC II.4.A and Initial Parameters Input II.4.B and Initial Parameters Input II.4.C and Initial Parameters Input All Various Partially Conforms Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.5.1-15.5.2, DSRS 15.5.1-15.5.2, Rev 0: Chemical and Volume Control System Malfunction That IncreasesReactor Coolant Inventory SRP 15.6.1, Rev 2: Inadvertent Opening of a PWR Pressurizer Reliefa BWR Pressure Valve or Valve Relief DSRS 15.5.1-15.5.2, DSRS 15.5.1-15.5.2, Rev 0: Chemical and Volume Control System Malfunction That IncreasesReactor Coolant Inventory DSRS 15.5.1-15.5.2, Rev 0: Chemical and Volume Control System Malfunction That IncreasesReactor Coolant Inventory
Tier 2 1.9-176 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.0.3 15.6.2 15.0.3 15.6.2 15.0.3 15.6.3 Not Applicable of this acceptance Comments Section The part of this guidance specifying the guidanceThe part specifying of this of consequences radiological of calculation containment of failure outside a postulated coolanta small reactor lineis applicableto the DCA. However, per SRP Section 15.0.3, I, Areas of Review,Section under Item 10.E Review Interfaces,part thesubheading of specifies this acceptance criterion that acceptance criteria is radiological Section 15.0.3. superseded by SRP The partguidance of this related to the specification for primary required technical is applicable to the activity iodine coolant However, per SRP Section 15.0.3, DCA. I, Areas of Review,Section under Item 10.E Review Interfaces,part thesubheading of specifies this acceptance criterion that acceptance criteria is radiological Section 15.0.3. superseded by SRP the guidanceThe part specifying of this of consequences radiological of calculation is failure steam generator tube a postulated per SRP However, to the DCA. applicable I, Areas of Review, 15.0.3, Section Section under subheading ReviewItem 10.F Interfaces, the part criterion that specifies radiological byacceptancecriteria is superseded SRP Section 15.0.3. acceptance criterion specifiesThis radiological analysis methodology and SRP that are supersededassumptions by Section 15.0.3. Status Review Plan (SRP) and Design Specific Review Not Applicable Partially Conforms Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Penultimate paragraph of SectionPenultimate II - Acceptability of Site on page 15.6.22 andESF Systems Dose Mitigating (2) and (1) Items and paragraph First of Section II on page 15.6.32 - Acceptability of Site and Dose Systems ESF Mitigating 15.6.22 - Plant-Specific Technical 15.6.22 Specifications for Primary Coolant System Iodine Activity - Section II on page 15.6.32 Methodology and Assumptions for Calculating Radiological Consequences AC II (No Number) II (No Number) Last paragraphSection II on page of II (No Number) II (No Number) First sentencelast of the paragraph of Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.6.2, Rev 2: Radiological Rev 2: Radiological SRP 15.6.2, of the Failure Consequences of Small Lines Carrying Primary Coolant Outside Containment Rev 2: Radiological SRP 15.6.2, of the Failure Consequences of Small Lines Carrying Primary Coolant Outside Containment Rev 2: Radiological SRP 15.6.3, Consequences of Steam Generator Tube Failure (PWR) Rev 2: Radiological SRP 15.6.3, Consequences of Steam Generator Tube Failure (PWR)
Tier 2 1.9-177 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.0.3 15.6.3 15.6.5 15.6.5 Not Applicable ns, SRP Section 15.0.3 ns, SRP technically relevant of this acceptance Comments Section The partguidance of this related to the specification for primary required technical secondary coolantand iodine activityis per SRP However, to the DCA. applicable I, Areas of Review, 15.0.3, Section Section under subheading ReviewItem 10.F Interfaces, the part criterion that specifies radiological byacceptancecriteria is superseded SRP Section 15.0.3. criteriasection and its acceptance SRP This are applicableonly to BWRs. design supports an exemptionThe NuScale 10selected portions of from CFR 50 Appendix K of K. The features Appendix requirements that are the NuScale design are included in to Appendix K analysissmall of break LOCAs. of Areas Section I, Section 15.0.3, SRP Per Review, under subheading Item 10 Review for the review of designInterfaces, certification applicatio radiologicalsupersedes the analyses, assumptions,acceptance criteria, and methodologies identified in this SRP Section 15.6.5. Status Review Plan (SRP) and Design Specific Review Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Evaluation of ECCS Performance of ECCS Evaluation Departure Radiological ConsequencesMost of Severe LOCA Plan Requirements Action TMI Conforms None. 15.6.5 paragraph of Section II on page - Plant-Specific Technical 15.6.32 Specifications for Primary and Iodine Coolant System Secondary Activity AC II (No Number)of the last Last two sentences All - Applicable Not II.1 II.2 II.3 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.6.3, Rev 2: Radiological Rev 2: Radiological SRP 15.6.3, Consequences of Steam Generator Tube Failure (PWR) Rev 2: Radiological SRP 15.6.4, of Main Steam Consequences Line Failure Outside Containment (BWR) DSRS 15.6.5, Rev 0: Loss-of- Coolant Accidents Resulting of Postulated from Spectrum Breaks within the Piping Pressure Reactor Coolant Boundary DSRS 15.6.5, Rev 0: Loss-of- Coolant Accidents Resulting of Postulated from Spectrum Breaks within the Piping Pressure Reactor Coolant Boundary DSRS 15.6.5, Rev 0: Loss-of- Coolant Accidents Resulting of Postulated from Spectrum Breaks within the Piping Pressure Reactor Coolant Boundary
Tier 2 1.9-178 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.0.3 15.6.5 15.0.3 15.6.5 15.6.5 pplicable only to BWRs. Comments Section The part of this guidance specifying the guidanceThe part specifying of this of consequences radiological of calculation a hypotheticalthe is applicable to LOCA However, per SRP Section 15.0.3, DCA. I, Areas of Review,Section under Item 10.H Review Interfaces,part thesubheading of specifies this acceptance criterion that acceptance criteria is radiological Section 15.0.3. A portion superseded by SRP of this guidance is a the guidanceThe part specifying of this calculationcontainmentof post LOCA contributionleakage is applicablethe to However, per SRP Section 15.0.3, DCA. I, Areas of Review,Section under Item 10.H Review Interfaces,part thesubheading of this Acceptancethat specifies Criterion radiological acceptance criteria and analysis Section 15.0.3. SRP model is superseded by applicableis this guidance only A portion of to BWRs. Status Review Plan (SRP) and Design Specific Review Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance ESF System Leakage AssumptionsESF System Leakage Conforms None. 15.0.3 Leakage Contribution LeakageLOCA Containment Contribution AC II.4 Programmatic RequirementsII.1 Conforms Calculatedand Doses Containment None.II.2 Post- of Calculation and for Model II.1 15.6.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.6.5, Rev 0: Loss-of- Coolant Accidents Resulting of Postulated from Spectrum Breaks within the Piping Pressure Reactor Coolant Boundary Rev 1: 15.6.5.A, SRP Radiological Consequences of a Design Basis Loss-of- Coolant Accident Including Containment Leakage Contribution Rev 1: 15.6.5.A, SRP Radiological Consequences of a Design Basis Loss-of- Coolant Accident Including Containment Leakage Contribution 15.6.5.B, Rev 1: SRP Radiological Consequences of a Design Basis Loss-of- Coolant Accident: Leakage From Engineered Safety Components Feature Outside Containment
Tier 2 1.9-179 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.0.3 15.6.5 15.0.3 15.6.5 15.6.6 Not Applicable pplicable only to BWRs. Comments Section le evaluated CHF as it is more is CHF as it le evaluated The part of this guidance specifying the guidanceThe part specifying of this of consequences radiological of calculation postulated leakage is applicable to the DCA. I, 15.0.3, Section However, per SRP Section Areas of Review, Item 10.H under Review Interfaces,part thesubheading of specifies this acceptance criterion that radiological analyses, assumptions, acceptancecriteria, and methodologies is Section 15.0.3. superseded by SRP that guidanceThe part specifying of this radiological consequences from ESF component leakage should be combined with consequences from other fission pathsproduct release is applicable to the However, per SRP Section 15.0.3, DCA. I, Areas of Review,Section under Item 10.H Review Interfaces,part thesubheading of specifies this acceptance criterion that acceptance criteria is radiological Section 15.0.3. A portion superseded by SRP of this guidance is a criteriasection and its acceptance SRP This are applicableonly to BWRs. appropriate than DNBR for the NuScale design. Status Review Plan (SRP) and Design Specific Review Conforms None. 15.6.6 Partially Conforms Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Consequences Consequences design value. AC II.2 Calculation of Radiological II.3 Radiological Combining All VariousII.1II.2 RCS pressure below 110 percent DNBR. Maintain minimum Not Applicable Conforms NuSca Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.6.5.B, Rev 1: SRP Radiological Consequences of a Design Basis Loss-of- Coolant Accident: Leakage From Engineered Safety Components Feature Outside Containment 15.6.5.B, Rev 1: SRP Radiological Consequences of a Design Basis Loss-of- Coolant Accident: Leakage From Engineered Safety Components Feature Outside Containment SRP 15.6.5.D, Rev 1: Radiological Consequences of a Design Basis Loss-of- Coolant Accident: Leakage From Main Steam Isolation Valve Leakage Control System (BWR) DSRS 15.6.6, Rev 0: Inadvertent Operation of ECCS DSRS 15.6.6, Rev 0: Inadvertent Operation of ECCS
Tier 2 1.9-180 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.2 15.7.4 Not Applicable Not Applicable ndling systems inside Comments Section Branch Technical Positionis 11-6, which referenced in Section 11.2. This acceptance criterion specifiesThis radiological analysis acceptance criteria 15.0.3. that are superseded by SRP Section criterionof this acceptanceThe portion related to fuel storage and handling systemsthe Fuel inside Building is applicableto those systems inside the NuScale Reactor Building.portion The of this acceptance criterion related to fuel storage and handling systems inside containment is applicable only to large LWR containment a incorporate that designs buildingplant housing numerous SSC. The use a containment does not NuScale design itsbuilding. own each NPM has Rather, compact steel containment vessel. This not containfuel containment vessel does storage and handling systems. Thus, the portion of this acceptance criterion related fuel storage and ha to applicable.containment is not acceptance criterion specifiesThis radiological analysis methodology and SRP that are supersededassumptions by Section 15.0.3. Status Review Plan (SRP) and Design Specific Review Conforms None. 15.6.6 Not Applicable Not Applicable Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance Acceptability of Site and Dose Acceptability of Site and Dose Systems ESF Mitigating serious plant condition without other other serious plant condition without faults occurring independently. Handling Systems Storage and Assumptions AC II.3Alldevelop more An AOO should not II.1 VariousII.2 Features of Fuel Radioactivity Control Partially Conformsto relocated The technicalbeen content has II.3 Modeling and Dose Model Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.6.6, Rev 0: Inadvertent Operation of ECCS 15.7.3, Rev 2: Radioactive SRP Release from a Subsystem or Component Rev 1: Radiological SRP 15.7.4, Consequences of Fuel Handling Accidents Rev 1: Radiological SRP 15.7.4, Consequences of Fuel Handling Accidents Rev 1: Radiological SRP 15.7.4, Consequences of Fuel Handling Accidents
Tier 2 1.9-181 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.7.4 Not Applicable Comments Section The NuScale design does not rely on ESF on The NuScale design does not rely mitigatethe ventilation systems to consequences of a design basis accident. Non-safety-related normal ventilation systemsatmosphere provide cleanup capability,necessary, as that meets the design, testing, and maintenance These systems 1.140. in RG guidelines containment,provide appropriate confinement, and filtering to limit releases of airborne radioactivity to the environment normal operations,during anticipated operational occurrences, and postulated accident conditions. However, these systems are not requiredfollowingan accident, and receive no credit in the determination of the radiological consequences of an accident. isof this acceptance criterion The intent but the specific languagerefers applicable LWR designs that incorporateto a containment buildingfuel within which performed.handling operations The are use a containment does not NuScale design itsbuilding. own each NPM has Rather, compact steel containment vessel immediately surrounding the reactor vessel. The containment design provisions of this fuel handling operationsguidance for inside containment are not relevant to the NuScale containment vessel design. this acceptance intent of the However, to the apply is appropriate to criterion NuScale Reactor Building, where the resideoperating NPMs in the reactor pool fuel handling operations are and performed. Status Review Plan (SRP) and Design Specific Review Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance ESF Grade Atmosphere Clean-Up Clean-Up Atmosphere ESF Grade Spentin Fuel Storage Area System AC II.4 II.5 Radiation Detection in Containment PartiallyConforms Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.7.4, Rev 1: Radiological Rev 1: Radiological SRP 15.7.4, Consequences of Fuel Handling Accidents Rev 1: Radiological SRP 15.7.4, Consequences of Fuel Handling Accidents
Tier 2 1.9-182 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.7.5 15.7.6 Not Applicable Not Applicable ea. The RBC system ea. The RBC Comments Section design conforms to the design conforms to the design conforms to the design conforms to the single-failure-proof guidelines of NUREG- single-failure-proof any credible0612 so that 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 evaluationsneed to perform load drop and as a result no accident analysisbeen has performed to assess radiological consequences of a spent fuel cask drop accident.accident or a NPM drop single-failure-proof guidelines of NUREG- single-failure-proof any credible0612 so that 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 evaluationsneed to perform load drop and as a result no accident analysisbeen has performed to assess radiological consequences of a spent fuel cask drop accident.accident or a NPM drop One of the principal functions of the craneto (RBC) isNuScale reactor building casks in the Reactormove spent fuel ar Building refueling the single-failure-proof conforms to design so that any of NUREG-0612 guidelines ofcredible a single failure component will stop to capability of loss the in result not and hold a critical load. The single-failure- proof crane precludes the need to perform loadevaluations dropno and as a result accident analysis has been performed to assess radiological consequences of a spent fuel caskaccident dropdrop or a NPM accident. Status Review Plan (SRP) and Design Specific Review Not ApplicableNot RBC system The Not ApplicableNot RBC system The Standard (DSRS) (Continued) AC Title/Description Conformance Acceptability of Site and Dose Acceptability of Site and Dose Systems ESF Mitigating Storage and Handling Systems Storage and AC All VariousII.1 Partially Conforms II.2 Features of Fuel Radioactivity Control Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.7.5, Rev 2: Spent Fuel 15.7.5, Rev 2: Spent SRP Cask Drop Accidents Fuel 15.7.5, Rev 2: Spent SRP Cask Drop Accidents Fuel 15.7.5, Rev 2: Spent SRP Cask Drop Accidents
Tier 2 1.9-183 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.7.5 15.7.6 Not Applicable Not Applicable Not Applicable Comments Section design conforms to the design conforms to the design conforms to the design conforms to the em design conforms to the em design conforms to the single-failure-proof guidelines of NUREG- single-failure-proof any credible0612 so that 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 evaluationsneed to perform load drop and as a result no accident analysisbeen has performed to assess radiological consequences of a spent fuel cask drop accident.accident or a NPM drop plant (See the acceptancecriteria in II.3). single-failure-proof guidelines of NUREG- single-failure-proof any credible0612 so that 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 evaluationsneed to perform load drop and as a result no accident analysisbeen has performed to assess radiological consequences of a spent fuel cask drop accident.accident or a NPM drop This guidance isguidanceThis only applicable to BWRs. Not Applicable single-failure-proof guidelines of NUREG- single-failure-proof any credible0612 so that 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 evaluationsneed to perform load drop and as a result no accident analysisbeen has performed to assess radiological consequences of a spent fuel cask drop accident.accident or a NPM drop Status Review Plan (SRP) and Design Specific Review Not ApplicableNot RBC system The Not Applicable Not Applicablecharacterized NuScale isevolutionaryan as Not ApplicableNot RBC system The Partially ConformsPartially The RBC syst Standard (DSRS) (Continued) AC Title/Description Conformance ESF Grade Atmosphere Clean-Up Clean-Up Atmosphere ESF Grade Spentin Fuel Storage Area System Criteria for Boiling Water Acceptance Reactors (BWRs) AcceptanceCriteria for Pressurized (PWRs) Water Reactors Assumptions Need for Calculation AC II.3 Modeling and Dose Model II.4 II.5 Eliminating Design Features Plant II.1 II.2 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.7.5, Rev 2: Spent Fuel 15.7.5, Rev 2: Spent SRP Cask Drop Accidents Fuel 15.7.5, Rev 2: Spent SRP Cask Drop Accidents Fuel 15.7.5, Rev 2: Spent SRP Cask Drop Accidents 15.8, Rev 2: Anticipated SRP Scram without Transients 15.8, Rev 2: Anticipated SRP Scram without Transients
Tier 2 1.9-184 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.8 15.8 15.8 Not Applicable Not Applicable e comment abovee for Comments Section The NuScale design relies on diversity 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 elements of the ATWS the diverse scram Rule. Asdiscussed in th AcceptanceII.3.A.i, the NuScale Criteria to RPS the within diversity on relies design reduce the risk associated with ATWS events. As discussed above in the comment for Acceptance CriteriaII.2, the design features required by 10 CFReither do not 50.62(C)(1) apply to the NuScale design or are not from ATWS the risk to reduce required events. Internal diversity within the MPS is a approachto addressing the diverse simpler ATWSscram elements of the Rule and criteria II.3.A.ii. and II.3.C(2) foracceptance evolutionary plants. NuScale conforms to the second criterion option of reducing the probabilitya of This is achievedto scram.failure with a insteadscram RPS of a diverse diverse system as discussed above. CHF preventReactor violation of trip signals limits before flow instabilities can develop. Status Review Plan (SRP) and Design Specific Review Conforms Not Applicable Not Applicable Partially Conforms Standard (DSRS) (Continued) iling the ATWS AC Title/Description Conformance Success Criteria is Sufficiently Small Success or, Demonstrate that the ATWS event consequences are acceptable the Failure Analysis Demonstrating Probability of Fa GDC 12Meeting Requirements of Conforms None. 4.4.7 to Design for demonstrating acceptable for demonstrating consequencesstability of AC II.3.A.iscram system Provide a diverse Conforms Partially II.3.A.ii II.3.B Apply Not Does Equipment Required II.3.C II.1II.2 No requirements - None. - II.3and suppress system criteria Detect Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 15.8, Rev 2: Anticipated SRP Scram without Transients 15.8, Rev 2: Anticipated SRP Scram without Transients 15.8, Rev 2: Anticipated SRP Scram without Transients 15.8, Rev 2: Anticipated SRP Scram without Transients DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities
Tier 2 1.9-185 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.2 4.4.4 4.4.7 4.4.7 4.4.7 4.4.7 4.4.6 4.4.7 15.9.A 15.9.A Not Applicable Not Applicable Not Applicable ns that could initiate ns before unstable flow rols on allowable allowable on rols reactor prior to conditions Comments Section Exclusion zone option is not used in theExclusion prevent signals NuScale design. Reactor trip violation of CHF limits can develop. Protectiveoscillations action occurs prior to development of oscillation. trip signals provide an reactor Existing exclusion zone that preventsviolation of from other causes, which is limits SAFDL morelimiting thanalready the exclusion instabilities. zone needed to preclude flow Detectoptions and Suppress are not Existing technical specifications employed. RTSfor provide cont unavailabilities of protectiveBackuptrips. options are not required. system trips reactor prior to conditions RTS initiatethatinstabilities. could flow Stabilities are not detected and suppressed. initiatethatinstabilities. could flow Stabilities are not detected and suppressed. CHF preventReactor violation of trip signals limits before flow instabilities can develop. detect unique monitoring is required to No hydraulic instabilities. RTS D&S. a of instead used is system RTS conditio occurs prior to instabilities. Status Review Plan (SRP) and Design Specific Review Conforms None. 4.4.4 Conforms Not Applicable Not Applicable ApplicableNot trips system RTS Partially Conforms Partially Conforms Partially Conforms Standard (DSRS) (Continued) at could violate AC Title/Description Conformance Detect and Suppress Method Trip of of Trip Method Suppress and Detect reactor before SAFDL violation Criteria to determine the acceptability of the D&S System compliancethe requirements of with GDC 20 free from other is plant Ensure instability modes th SAFDLs D&S System extremely high probability of functioning in the event of an AOO. Exclusion zoneand buffer region methodology declared inoperable monitor process variables and systems. should be functionality demonstrated by analysis. AC II.4and Suppress Detect Method: II.5 II.6 Backupif licensing options solutions II.7 II.8II.9and Detect Suppress system to II.10 Stability-related instrumentation II.11 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities DSRS 15.9.A, Rev 0: Thermal Review Hydraulic Stability Responsibilities
Tier 2 1.9-186 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Ch 16 Consistentthis with specifications that is er, the DCA contains ch of the specific ch of the Comments Section This DSRS section and its acceptance criteria section and its acceptanceThis DSRS is applicable but mu technical LWR to existing refers language specifications or to plant-specific technical specifications to bedeveloped by a COL applicant. For the latt COL information items, as appropriate, that of describe therequired development plant-specific technical COL applicant referencing to the deferred the Notwithstanding design. NuScale the above,pursuant to 10 CFR and 52.47(a)(11) consistent with DSRS 16.0, the DCA contains proposedtechnical specifications that are prepared in accordance with 10 CFR 50.36 and 10 CFR The 50.36a. improved standard technical specification guidance for LWRs - NUREGs-1430 DSRS specified in this - were and NUREG-2194 through -1434, appropriatethe extent utilized to and practicable. Additionally, the Technical “Writer’s Guide for Specifications Task Force Plant-Specific Improved Technical Revision 1, Specifications,” TSTF-GG-05-01, August 2010 was used todraft the specifications. technicalThere are a number of and differenceseditorial between the NuScale proposed technical specifications and those presented in the improved standard technical specifications. justification for such technical 16.0, DSRS provided. is differences Status Review Plan (SRP) and Design Specific Review Partially Conforms Standard (DSRS) (Continued) AC Title/Description Conformance AcceptanceCriteria for Technical Specifications AC All (No Number) Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: DSRS 16.0, Rev. 0: Technical DSRS Specifications
Tier 2 1.9-187 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 16.1.1 16.1.1 Not Applicable Not Applicable r than SRP Section Section r than SRP able only to existing able only to existing not applicable to not applicable to s to revisionss to being Comments Section This guidance is for revisions being made to existing technical specifications (TS), presumably including deviation from generic or any applicable standard TS. The was considered in the discussion provided development of the NuScale TS, however is the specific content of newas a part genericdevelopment of TS a DCA. This guidance applie made to existing TS, including deviation generic or applicable standard TS. The from was considered in the discussion provided development of the NuScale TS, however is the specific content of newas a part genericdevelopment of TS a DCA. This guidance is applic NRC-approvedthat are based QA Programs and its daughter standards. on ANSI N45.2 The NuScale QA Program Description and the (QAPD) is based on NQA-1-2008 endorsed in RG addenda, as NQA-1a-2009 4. Since the issuance of SRP Rev 1.28, NRC has issued SRP 17.1, the Section on NQA-1) for the 17.5 (based Section review of QAPDs for new reactor applicants including applicants for design - certification - under 10 CFR 52. Accordingly, 17.5 (rathe SRP Section 17.1) is the appropriate guidance to be applied to the NuScale QAPD. This guidance is applic NRC-approved operational QA Programs that are based on ANSI N45.2 and its daughter standards. Status Review Plan (SRP) and Design Specific Review Guidelines Partially Conforms Standard (DSRS) (Continued) ines Partially Conforms AC Title/Description Conformance AC II.1 Engineering Traditional II.2 Probabilistic Guidel All VariousAll Not Applicable Various Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 16.1, Rev 1: Risk-Informed SRP 16.1, Decision Making: Technical Specifications Rev 1: Risk-Informed SRP 16.1, Decision Making: Technical Specifications 17.1, Rev 2: Quality SRP the Design Assurance During and Construction Phases 17.2, Rev 2: Quality SRP the Assurance During Operations Phase
Tier 2 1.9-188 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 17.5 17.5 Not Applicable Not Applicable rdingly, SRP Section able only to existing Comments Section This guidance is applic the SinceNRC-approved QA Programs. has NRC section, the issuance of this SRP review of for the Section 17.5 issued SRP QAPDs for new reactor applicants - includingfor design applicants certification - under 10 CFR 52. Acco is the Section 17.3) than SRP (rather17.5 guidanceappropriate to be applied to the QAPD incorporated into the DCA. acceptance criterion is applicable only This to COL applicants. The onsite, offsite, operational, and maintenance organizationalof elements COLItem II.A.3 are the responsibility of the certifiedapplicant referencing design. for site-specific and The provisions operationalof phasethe quality assurance applicableprogram are not NuScaleto the QA program to be applied during the design certification phase, and are to be QA operationaladdressed within the program developed and maintained by the applicantcertified referencing theCOL design. Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance COL Applicant Not Applicable AC All VariousII.AII.B II.A Design Certification Not Applicable OrganizationII.B Conforms Quality Assurance Program None. Partially Conforms PartiallyConforms II.C Design Controland Verification Conforms None. 17.4 17.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 17.3, Rev 0: Quality SRP Assurance Program Description 17.4, Rev 1: Reliability SRP (RAP) Program Assurance 17.4, Rev 1: Reliability SRP (RAP) Program Assurance 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and
Tier 2 1.9-189 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 17.5 17.5 Not Applicable Not Applicable Comments Section The site-specific and operational provisions controldocument are the responsibilityof the referencing the COL applicant of certified design. acceptance criterion governs activities This responsibilitythat are the of the COL the certifiedapplicant referencing design. acceptance criterion governs activities This responsibilitythat are the of the COL the certifiedapplicant referencing design. inservice, specific to The provisions modification,responsibilityetc. are the of the COL applicant referencingcertified design. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 17.5 17.5 Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Drawings (Controlled Documents) Services and Equipment, Parts, and Components Materials, AC II.D Procurement Document ControlII.E Conforms Instructions, Procedures, and II.F None. Document ControlII.GPurchased Material, Control of II.H PartiallyConforms Identificationand Control of II.I 17.5 ControlSpecial of ProcessesII.J Not Applicable Inspection Partially Conforms Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and
Tier 2 1.9-190 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Comments Section This acceptance criterion governs activities This responsibilitythat are the of the COL the certifiedapplicant referencing design. acceptance criterion governs activities This responsibilitythat are the of the COL the certifiedapplicant referencing design. Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 17.5 17.5 Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Equipment Status Components AC II.K Test ControlII.LMeasuring and Test of Control II.M Handling, Storage,and Shipping ConformsII.N None. Not Applicable Inspection, Test, and Operating II.O Nonconforming Materials, Parts, or II.P Corrective ActionII.Q 17.5 Quality Assurance Records Conforms Conforms None. None. 17.5 17.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and
Tier 2 1.9-191 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable ion is applicable to Comments Section This SRP section and its acceptance criteriasection and its acceptance SRP This govern a site-specific operational program that is the responsibilitythe COL of applicant. This acceptance criter existing reactor licensees that request NRC approval modifications. of control room Status Review Plan (SRP) and Design Specific Review Conforms None.Conforms None. 17.5 18.1 thru 18.12 Conforms None.Conforms None. 17.5 17.5 Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Nonsafety-Related SSC Quality Quality SSC Nonsafety-Related Controls New a HFE Aspects of Review of the Plant Control HFE Aspects of Review of the Modifications Room Inspection and Test Inspection and Commitments AC II.R AuditsII.S Training and Qualification CriteriaII.T Conforms Training and Qualification - None.II.U Conforms None.II.V Quality Assurance Program AllII.A Various 17.5 II.B 17.5 Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.5, Rev 1: Quality SRP Assurance Program Design Description - Certification, Early Site Permit applicants COL New and 17.6, Rev 2: Maintenance SRP Rule 18.0, Rev 2: Human SRP Factors Engineering 18.0, Rev 2: Human SRP Factors Engineering
Tier 2 1.9-192 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 18.4 18.6 18.4 18.6 18.4 18.6 19.0 19.1 19.2 18.4 18.6 7.1.5 7.1.5 7.1.5 7.1.5 Not Applicable Not Applicable ion is applicable to Comments Section Section 3, "Crediting Manual Operator ActionsDefense-in-Depth in Diversity and Analyses”. (D3) Section 3, "Crediting Manual Operator ActionsDefense-in-Depth in Diversity and Analyses”. (D3) Section 3, "Crediting Manual Operator ActionsDefense-in-Depth in Diversity and Analyses”. (D3) This acceptance criter existing reactor licensees that request NRC approval of plant changesthat affect important human actions. Evaluationsite-specificPRA of hazards and update are COL applicant responsibility. a licensee to by used to PRAs Applicable amendmentssupport license for an operating reactor. Section 3, "Crediting Manual Operator ActionsDefense-in-Depth in Diversity and Analyses”. (D3) ormsappendix This supersedes DI&C ISG-05, Status Review Plan (SRP) and Design Specific Review Conformsappendix This supersedes DI&C ISG-05, Conformsappendix This supersedes DI&C ISG-05, Conformsappendix This supersedes DI&C ISG-05, Not Applicable Standard (DSRS) (Continued) AC Title/Description Conformance Review of the HFE Aspects of Review of the Risk Affecting Modifications Actions Human Important (Preliminary Validation) Validation) System (Integrated (Maintaining Long-Term Integrity of Actions in the D3 Credited Manual Analysis) AC II.C C.1.B (Analysis)Phase 1 Review Criteria for C.2.B Conf Phase 2 Review Criteria for C.3.B Phase 3 Review Criteria for C.4.B Phase 3 Review Criteria for AllAll Various Various Partially Conforms Not Applicable Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 18.0, Rev 2: Human 18.0, Rev 2: Human SRP Factors Engineering SRP Appendix 18-A, Rev 0: Crediting Manual Operator and in Diversity Actions Defense-in-Depth (D3) Analyses SRP Appendix 18-A, Rev 0: Crediting Manual Operator and in Diversity Actions Defense-in-Depth (D3) Analyses SRP Appendix 18-A, Rev 0: Crediting Manual Operator and in Diversity Actions Defense-in-Depth (D3) Analyses SRP Appendix 18-A, Rev 0: Crediting Manual Operator and in Diversity Actions Defense-in-Depth (D3) Analyses 19.0, Rev 3: Probabilistic SRP Risk Assessment and Severe for New Evaluation Accident Reactors 19.1, Rev 3: Determining SRP The Technical Adequacy of Probabilistic Risk Assessment License For Risk-Informed Amendment Requests After Initial Fuel Load
Tier 2 1.9-193 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 20 19.4 Not Applicable ed by NEI 06-12 and 06-12 ed by NEI Comments Section proposalsfor changes to thelicensingbasis. Applicable with theexception of acceptance criterion II.17 Boiling Water Reactor: Containment Venting and Vessel Flooding (Item B.2.e) which is a BWR specific criterion and acceptance criterion II.20 SFP Mitigative Measures. The SFP mitigating requir measure is not a statement that this mitigation includes is below required if the SFP is not strategy grade and cannot be drained. TheNuScale SFP is below grade and cannot be drained. Status Review Plan (SRP) and Design Specific Review Standard (DSRS) (Continued) AC Title/Description Conformance AC All VariousAllAll Various Various Not Applicable Applicable to licensees, plant-specific All Conforms None. Partially Conforms Various Conforms None. 19.3 19.5 Table 1.9-3: Conformance with NUREG-0800, Standard with NUREG-0800, Conformance Table 1.9-3: Title SRP or DSRS Section, Rev: SRP 19.2, Initial Issuance: Issuance: 19.2, Initial SRP of Risk Information Review Permanent Used to Support PlantSpecificChanges to the GeneralLicensing Basis: Guidance 19.3, Rev 0: Regulatory SRP Treatment of Non-Safety Passive Advanced Systems for Light Water Reactors 19.4, Rev 0: Strategies SRP and Guidance to Address Loss-of-Large Areas of the to Explosions and Due Plant Fires of 19.5, Rev 0: Adequacy SRP features and Design functional capabilities for described and identified Impacts Aircraft withstanding
Tier 2 1.9-194 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 3.7 3.7 2.5 Not Applicable Not Applicable rk activities ion 3.7 are consistent.ion 3.7 rform limited wo 1/3 of the CSDRS thus does all ESP and all ESP and COL applicants Comments Section finitions providedfinitions in Sect Sections 2, 3, 4, and 5. 2, 3, 4, Sections The CSDRS (and CSDRs-HF) is effectively the SSE for the for the the SSE is effectively (and CSDRs-HF) The CSDRS The OBE is specified as DCA. COL items arethe DCA. Thereinnot require any analysis and to GMRS is enveloped for the applicant to ensure have a seismic monitoring program with responses following anOBE exceedance. highfrequency CSDRS.The NuScale design includes a Thislaboratory discusses analysisthe site-specific of soil 3.7 COLthe applicant items forThe FSAR includes column. to develop site-specific information. applicableisThisapplicants. ISG to COL reviewPRA information and concerningGuidance the of and severe accident assessments submitted to support DC Rev COL applications has been 19.0, incorporated into SRP Not Applicable 3. This ISG is applicable to requesting authorization to pe or considering preconstruction activities. -in the guidance provided This sectionpoints out to Status Conforms Conforms Partially Conforms nts onnts the Industry AC Title / Description Conformance Table 1.9-4: Conformance with Interim Staff Guidance (ISG) Interim Staff Guidance Draft White Paper on Testing of for Dynamic Soil Properties Nuclear Power Plant Combined License Applications and on Information for Guidance Review Staff Guidance/Position on the Staff Guidance/Position DefinitionsSafe-Shutdown of and Operating-Basis Earthquakes, Use of Various Seismic Motions, Ground Operating- Instrumentation and Basis EarthquakeExceedance on Staff Guidance/Position Addressing HF Ground Motion Evaluations AC 1 Seismic Issues addressed in this AllAll Various VariousAll Various Not Applicable Not Applicable Not Applicable of Applicants ISG Section/Title For Combined For Combined License Applications DC/COL-ISG-1: Seismic DC/COL-ISG-1: Seismic High Frequency of Issues Ground Motion DC/COL-ISG-1DC/COL-ISG-1 2 3 DC/COL-ISG-1 Ground Motion DefinitionsDC/COL-ISG-1 4 Conforms The de 5DC/COL-ISG-2: Financial Qualifications Staff Comme DC/COL-ISG-3: Probabilistic Risk Assessment Information Design to Support Combined Certification and License Applications DC/COL-ISG-4: Definition of Construction and on Limited Work Authorizations
Tier 2 1.9-195 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 2.3 3.4 3.8 3.9.5 12.3.6 Not Applicable ssary because the existing ssary because is guidance applicable, design features, facilities, design features, calculations. However, an Comments Section imilar to large PWRs in the existing imilar to large PWRs pplicable except for aspects that are fleet for effluent release for effluent fleet is nece methodology alternate evaluation1980s for PWRGALE code was developed in the of the large PWR reactorsthat time and does not of address the NuScale plant design. BWR-specific. This guidance refers to Attachment C. The correct correct C. The refers to Attachment This guidance B. reference is Attachment Th site-specific, that relate to the portions except for aspectsthe COL operational that are the responsibility of referencingapplicantof the NuScale design. The aspects this guidancethat pertain to to relevant that are technically functions, and equipment the NuScale standard plant design are applicable to the DCA. Section 3.4 identifies parameter specifiedthe Extreme for and Normal winter precipitation events. These values are the structuralThe COL applicant analysis used in in 3.8. needs to determine site-specific informationto compare to the design parameters. That determination is performed in Section 2.3. This portion of the ISG is applicable only to COL applicants. Not Applicable Status Conforms Conforms None. Ch 16 Not Applicable The NuScale design is s Not Applicable Partially Conformsa This guidance is Partially Conforms AC Title / Description Conformance Final Interim Staff Guidance on of GALE86 3 - Acceptability Page PrecipitationEvents and their Loads Roof Live Resulting AdverseReview of Flow Effects Acceptance- Criteria RG 4.21 Compliance with paragraph under heading First Final Interim Staff Guidance, identification and specifying timing of resolution generic technical specification COL action items Second, third, and fourth Final paragraphs under heading Interim Staff Guidance, specifying compliance options for COL applicants Table 1.9-4: Conformance with Interim Staff Guidance (ISG) (Continued) (ISG) with Interim Staff Guidance Conformance Table 1.9-4: AC All Five paragraphs under heading thru Bullets 1 & 4) (p 3 6 AllExtreme Winter Normal and para 1 (p4) & 5) All Final1 - paragraph on Page ISG Section/Title DC/COL-ISG-5: GALE86 Code DC/COL-ISG-5: GALE86 Routine for Calculation of Radioactive Releases in Effluents Liquid Gaseous and Design to Support Combined Certification and License Applications DC/COL-ISG-6: Evaluation and Acceptance Criteria for 10 CFR to Support 20.1406 Design Certification and Combined License Applications DC/COL-ISG-7: Assessment of Normal and Extreme LoadsWinter Precipitation on the Roofs of Seismic I Category Structures DC/COL-ISG-8: Necessary of Plant-Specific Content Technical Specifications DC/COL-ISG-8 Review of DC/COL-ISG-10: (p. 4 para 2-4 to Address Evaluation Adverse in Flow Effects ThanEquipment Other Reactor Internals
Tier 2 1.9-196 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 11.2.3 11.2.3 11.2.3 11.2.3 11.2.2 Not Applicable Not Applicable Not Applicable except for aspects that are ch, this guidance is the n is explicitly directed towards the Comments Section review of combined license applications. This guidance is applicableThis guidance is applicableapplicants only to COL site permit. or early Site-specificaspects that are the responsibilitythe COL of applicant. Thisapplicable guidance is except for site-specificaspects that are the responsibilityCOL applicant. of the Site-specificaspectsresponsibility are the of the COL applicant. This acceptancecriterion governs site-specific calculations that are the responsibility of the COL applicant referencing the certified design. This acceptance criterion governs analysis modeling using and characteristics, site data, hydrogeological site-specific radiological analysis;su as responsibilityof the COL applicant referencing the certified design. Site-specificaspectsresponsibility are the of the COL applicant. Site-specificaspects (e.g., development and implementation of the ODCM) are the responsibility of the COL applicant. Status Not Applicable Not Applicable This acceptance criterio Not Applicable Partially Conforms Partially Conforms Partially Conforms Partially Conforms ure Scenarios and AC Title / Description Conformance Freeze Point;That Changes Should Not beConsidered for Deferral Radioactivity Releases (Including Attachment A) or Water Ground in Capabilities Surface Water Radioactivity Concentration Levels Combined License Reviews SRP Dose Acceptance Criteria Partially Conforms Acceptance Criteria Table 1.9-4: Conformance with Interim Staff Guidance (ISG) (Continued) (ISG) with Interim Staff Guidance Conformance Table 1.9-4: AC All1 Licensing-Basis Information Failure Mechanism and ISG Section/Title DC/COL-ISG-11: Finalizing DC/COL-ISG-11: Licensing-basis Information NUREG- DC/COL-ISG-13: Review Plan 0800 Standard Section 11.2 and Branch Technical Position 11-6 Assessing the Consequences ofAccidental an Release of Radioactive Materials from Tanks for Waste Liquid Combined License Applications Submitted under 10 CFR Part 52 DC/COL-ISG-13DC/COL-ISG-13DC/COL-ISG-13 2 3DC/COL-ISG-13 4 Mitigating Design Features Radioactive Source Term Partially Conforms 5DC/COL-ISG-13 Calculations of Transport DC/COL-ISG-13 Expos 6 DC/COL-ISG-13 7 8Tank Waste Specifications on Evaluation Findings for
Tier 2 1.9-197 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Ch 1 Not Applicable Not Applicable Not Applicable d strategies intended to to strategies intended d ort combined license (COL) ns 2.4.12 and 2.4.13, this and 2.4.13, ns 2.4.12 applicants. Not Applicable t applicable to design certification Comments Section applicants; however 10 CFR 52.80(d) requires COL requires COL applicants; 52.80(d) however 10 CFR applicants to include a description and plans for the guide an implementation of core cooling, containment,restore maintainor and SFP cooling capabilities under the circumstances associated withdue to explosionsthe plant the LOLA of or fireas (hh)(2). 50.54 required by 10 CFR As a supplement to SRP Sectio to SRP As a supplement governsguidance site-specific hydrogeologicaldata, site characteristics, and radiologicalanalysis aspects that are the responsibility of the COL applicant referencingthe certified design. for COL This guidance is the DCA apply to that this guidance of The portions include discussion concerning COL actionitems and COL holder term “COL using the items and not information actionitems are identified throughout the item.” COL FSAR. applicableof seismic designis review to theThis ISG information submitted to supp applications. Status Not Applicable Not Applicable Partially Conforms AC Title / Description Conformance Interfaces; Regulatory Requirements; Onsite Hydrogeological Characterization; Contamination Source and Receptor Location; Groundwater Modeling and Pathway Prediction;and Radioactive Consequence Analysis New Section C.III.4.3 to Replace to Section C.III.4.3 New 1.206 RG of C.III.4.3 Section AnticipatedRevisions NRC of SRP Chapter NUREG0800, 1.0 Table 1.9-4: Conformance with Interim Staff Guidance (ISG) (Continued) (ISG) with Interim Staff Guidance Conformance Table 1.9-4: AC Allof Review; Review Area No Num (p4- 11) (p11-23) All -All - Not Applicable is no 10 CFR 50.54(hh)(2) Not Applicable ISG Section/Title DC/COL-ISG-14: Assessing Water Flow and Ground Transport of Accidental Releases Radionuclide Post- ESP/DC/COL-ISG-15: Combined License Commitments ESP/DC/COL-ISG-15 Num No Compliance DC/COL-ISG-16: with 10 CFR and 50.54(hh)(2) 10 CFR 52.80(d) Ensuring DC/COL-ISG-17: Hazard-Consistent Seismic and Site Response for Input Soil Structure Interaction Analyses
Tier 2 1.9-198 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 2.0 3.3 3.5 Not Applicable Not Applicable Not Applicable COL applicationshas been only to nuclearplants power instead of gas turbine instead gardless of the type of standby AC standby gardlessof the type of Comments Section circulation). The NuScale emergency establishes requirements for hurricane wind speed and missile spectra missile speed and wind hurricane Regulatory guidance“consistent in with requirements 1.221, R0.” Specific design and 3.5.1.4. established in Sections 3.3.2 are This guidance is applicableThis guidance is only to reactor plant designs for which operation of emergency core cooling, residual heat removal,containment and spray systems relies on pumps (i.e. forced NuScale heat removal systems (the coolingcoredecay and designinclude does not a containment spray system) operate via natural circulation, and do not require or include pumps. performanceconcerningGuidance the of a SMA DC and to support submitted Rev 3. 19.0, into SRP incorporated applicableThis guidance is that use a gas turbine-driven standby emergency AC - in lieupower system of emergency diesel generators - to supply power to safety-related or risk-significant equipmentduring postulated for operational events and accident conditions.onsiteNuScale The uses design backup diesel generators generators. However, re the onsite standby design, generationused in the NuScale distribution the onsite AC source and AC generation norserves are not safety-related, are they relied system it 72 hours the first during functions safety to fulfill upon following a design basis accident. applicableisThisapplicants. ISG to COL Not Applicable Status Not Applicable AC Title / Description Conformance Turbine Generators (Including 1) Attachment Table 1.9-4: Conformance with Interim Staff Guidance (ISG) (Continued) (ISG) with Interim Staff Guidance Conformance Table 1.9-4: AC All VariousAll Various Not Applicable All Guidance for Emergency Gas Not Applicable All VariousAll Various Not Applicable Conforms Section 2.0 ISG Section/Title DC/COL-ISG-19: Gas DC/COL-ISG-19: Accumulation Issues in Systems Related Safety DC/COL-ISG-20: Implementation of a Risk Probabilistic Assessment-Based Seismic Margin Analysis for New Reactors Review of DC/COL-ISG-21: Nuclear Power Plant Designs using a Gas Turbine Driven Standby Emergency Alternating Current Power System Impact of DC/COL-ISG-22: a (Under Construction of New License) Combined Nuclear Power Plant Units at Multi- on Operating Units Unit Sites DC/COL-ISG-24: Implementation of RG 1.221 Hurricane on Design-Basis and Hurricane Missiles
Tier 2 1.9-199 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 19.1 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable e applicants requesting a light water SMR reactor light the review of ESP andthereview COL Comments Section not applicable to the NuScale DCA. not applicable to the NuScale DCA. not applicable to the NuScale DCA. not applicable to the NuScale not applicable to the NuScale applicants to conform to PRA Standard. to PRA conform applicants to See DSRS in Table 7.1.5 which provides information 1.9-3 on the Diversity and Defense-in-Depth review. application. See DSRS 7.1.5 in Table 1.9-3 which provides informationDefense-in-Depthon Diversity and review. Tableinapplication. DSRS 7.1.5 See 1.9-3 which provides informationDefense-in-Depthon Diversity and review. See DSRS in Table 7.1.5 which provides information 1.9-3 on the Diversity and Defense-in-Depth review. See DSRS in Table 7.1.5 which provides information 1.9-3 on the Diversity and Defense-in-Depth review. This ISG is applicableto 10 CFR Part52,licensees COL with ChangesConstruction license during condition. This ISG is applicable to applications, including thos limited work authorization. OL, CP, LWA, ESP, of review applicable to the is This ISG for COL applications and technologies. Status Not ApplicableNot is Digital I&C-ISG-02 ApplicableNot is Digital I&C-ISG-02 ApplicableNot is Digital I&C-ISG-02 Not ApplicableNot is Digital I&C-ISG-02 ApplicableNot is Digital I&C-ISG-02 AC Title / Description Conformance Operator Actions - Staff Position Position - Staff Operator Actions (Pages 2 and 3) Position (Page 12) (Page 14) BTP 7-19 Position 4 Challenges - BTP 7-19 6) (Page Position Staff Cause Failure Effects of Common 8 Position (Pages - Staff (CCF) and 9) Table 1.9-4: Conformance with Interim Staff Guidance (ISG) (Continued) (ISG) with Interim Staff Guidance Conformance Table 1.9-4: AC AllAll VariousAll VariousAll Various Not Applicable Various Not Applicable 5. Not Applicable 1 and 2 Staff PositionDiversity and Manual Adequate Conforms and Provides guidance for DC COL Not Applicable Not Applicable ISG Section/Title DC/COL-ISG-25: Changes DC/COL-ISG-25: during Construction Under Code of the of Title 10 Regulations Part 52 Federal DC/COL-ISG-26: Environmental Issues New with Associated Reactors Specific DC/COL-ISG-27: for Environmental Guidance Water Small Modular Light Reactor DC/COL-ISG-28: Assessing Adequacythe Technical of the Advanced Light-Water Reactor Probabilistic Risk Assessment for the Design Certification Application and Combined License Application Cyber Digital I&C-ISG-01: Security Diversity Digital I&C-ISG-02: and Defense-in-Depth (D3) Digital I&C-ISG-02Digital I&C-ISG-02 3 Digital I&C-ISG-02 4 Digital I&C-ISG-02 6 7 Defense - Staff of Echelons Position Staff Single Failure -
Tier 2 1.9-200 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 18.7 18.7 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable "Design of the Highly the Highly of "Design Platform Topical Report." Platform Topical Report." Platform Topical Report." Platform Topical Report." Platform Topical Report." Platform Topical Report." except for site-specific site-specific except for Comments Section not applicable to the NuScale DCA; not applicable to the NuScale DCA; not applicable to the NuScale DCA; not applicable to the NuScale DCA; not applicable to the NuScale DCA; not applicable to the NuScale DCA; n subtier guidance. erion is applicable except for the except for erion is applicable 18 erion is superseded by Chapter is not applicable to the NuScale DCA. See DSRSTablein 7.0 1.9-3 which providesoverview an of the I&C review process. System Protection Integrated applicationcertai of 18-A. Appendix however, it is addressed as part of topicaladdressed report NuScale is as part of however, it Power, LLC, TR-1015-18653-P-A, topicaladdressed report NuScale is as part of however, it "Design Power, LLC, TR-1015-18653-A, of the Highly System Protection Integrated topicaladdressed report NuScale is as part of however, it "Design Power, LLC, TR-1015-18653-A, of the Highly System Protection Integrated topicaladdressed report NuScale is as part of however, it "Design Power, LLC, TR-1015-18653-A, of the Highly System Protection Integrated topicaladdressed report NuScale is as part of however, it "Design Power, LLC, TR-1015-18653-A, of the Highly System Protection Integrated topicaladdressed report NuScale is as part of however, it "Design Power, LLC, TR-1015-18653-A, of the Highly System Protection Integrated This position is applicable that are the subtier NUREG-0899 elements of operational responsibilityof the COL applicant. Status Not ApplicableNot is Digital I&C-ISG-04 ApplicableNot is Digital I&C-ISG-04 ApplicableNot is Digital I&C-ISG-04 ApplicableNot is Digital I&C-ISG-04 ApplicableNot is Digital I&C-ISG-04 ApplicableNot is Digital I&C-ISG-04 Not Applicable This acceptance crit Partially Conforms Partially Conforms This acceptance crit AC Title / Description Conformance - Staff Position (Pages - Staff Position 4 through 8) Position (Pages 8 through 10) Position Staff - Stations Display through 16) 11 (Pages through 13) 11 (Pages through 15) 13 (Pages (D3) Considerations (Page 15) 3 through (Pages Position Staff 7) Position (Pages 9 through 11) Actions in Diversity and Defense- In-Depth Analyses (D3) (Pages 21) through 13 Table 1.9-4: Conformance with Interim Staff Guidance (ISG) (Continued) (ISG) with Interim Staff Guidance Conformance Table 1.9-4: AC 41 Position Staff Interdivisional Communications Applicable Not Digital I&C-ISG-03 1 Computer-Based- Procedures ISG Section/Title Digital I&C-ISG-03: Risk- Digital I&C-ISG-03: Informed Digital Instrumentation and Controls Highly Digital I&C-ISG-04: Integrated Control Rooms & Digital Communication Systems Digital I&C-ISG-04 2Digital I&C-ISG-04 3Digital I&C-ISG-04 Prioritization - Staff Command 3.1Digital I&C-ISG-04 and Control Multidivisional Independence and Isolation 3.2Digital I&C-ISG-04 Factors Considerations Human 3.3 Highly Digital I&C-ISG-05: Integrated Control Rooms - Human Factors Digital I&C-ISG-05Defense-in-Depth Diversity and Digital I&C-ISG-05 2 3 Inventory Minimum - Staff Manual Operator Crediting
Tier 2 1.9-201 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable t used as items relied on facilities from unintentional facilities from requests for licensing basis requests for licensing ees to implement digital I&C Comments Section erion is superseded by Chapter 18 erion is superseded by Chapter 18 erion is superseded by Chapter 18 erion is superseded by Chapter 18 erion is superseded by Chapter license holder during decommissioning license holder during decommissioning Appendix 18-A Criterion 1.B. Appendix 18-A Criterion 2.B. Appendix 18-A.Criterion 3.B. Appendix 18-A.Criterion 4.B. activities. activities. This guidance is for review of This guidance is existingchanges licens from upgrades. measuresfor review of proposed for This guidance is protecting equipmen digital I&C cycle fuel at (IROFS) for safety digital events. site-specificThis guidance governs programmatic and design aspectsthat are the of emergency planning responsibilityof the COL applicant referencing the design. NuScale applicableisThisapplicants ISG to holders of, and for, licenses,operating constructionpermits, and combined licenses. Status Not Applicable This acceptance crit Not Applicable This acceptance crit Not Applicable This acceptance crit Not Applicable This acceptance crit AC Title / Description Conformance Criteria (Pages through 16) Criteria 15 (Page 18) Review Criteria Validation- Review Criteria through 20) 19 (Pages of Credited Manual Integrity ActionsAnalysis in the D3 - (Page 21) Review Criteria Table 1.9-4: Conformance with Interim Staff Guidance (ISG) (Continued) (ISG) with Interim Staff Guidance Conformance Table 1.9-4: AC AllAll VariousAll VariousAll Various Not Applicable VariousAll Not Applicable Various Not Applicable All Not Applicable Various Applicable to Not Applicable Applicable to Not Applicable ISG Section/Title Digital I&C-ISG-05Digital I&C-ISG-05 3.1.BDigital I&C-ISG-05 3.2.BDigital I&C-ISG-05 - Review 1: Analysis Phase 3.3.B 2: Preliminary Validation - Phase 3.4.BLicensing Digital I&C-ISG-06: System 3: Integrated Phase Process Fuel Digital I&C-ISG-07: Maintaining Phase 4: Long-Term Cycle Facilities NSIR/DPR-ISG-01: Emergency Planning for Nuclear Power Plants NSIR/DPR-ISG-02: Emergency Planning for Requests Exemption Decommissioning Nuclear Power Plants of Review NSIR/DPR-ISG-03: Security Exemptions/License Amendment Requests for Decommissioning Nuclear Power Plants 1: Rev JLD-ISG-12-01, Compliancewith Order EA- 12-049 Concerning Mitigation Strategies
Tier 2 1.9-202 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.1.1 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable e criticality analysis.e criticality 9.1.1 -ISG-020 remains the NRC’s remains -ISG-020 ation ofation anSMA to new credit for burnup. 9.1.1 Comments Section able to BWR licensees with Mark I and es not take credit for zoning or a loading es not take credit for zoning l analysis is appropriate for conditions.l analysis is appropriate for 9.1.1 mbly design is used in th used in is mbly design ysis does not take credit for burnup. 9.1.1 alysis does not take pattern. Mark II containments.Mark II This ISG is applicableisThisapplicants ISG to holders of, and for licenses,operating constructionpermits, and combined licenses. Poolcapable monitoring instrumentation that is design indication of beyond monitoring and providing of eventsbasismonitor (i.e., instrumentation that can wide a rangeof spent fuel levels) poolthe NuScale is part of design. request 50.54(f) March 2012 for response to the is This ISG for information letter. DC/COL applic for current guidance reactors licensing. request 50.54(f) March 2012 for response to the is This ISG for information letter. request 50.54(f) March 2012 for response to the is This ISG for information letter. The information in this guidance is site-specific and is the responsibilityof the COL applicant. Status AC Title / Description Conformance Table 1.9-4: Conformance with Interim Staff Guidance (ISG) (Continued) (ISG) with Interim Staff Guidance Conformance Table 1.9-4: AC All Various Not Applicableapplic is This ISG All VariousAll VariousAll Not Applicable All VariousAll Various Not Applicable Various Not Applicable Not Applicable Not Applicable ISG Section/Title DSS-ISG-2010-01 1Assembly Selection Fuel Conforms One asse fuel DSS-ISG-2010-01 2 Depletion Analysis Conforms The anal DSS-ISG-2010-01 3a Burnup Profile Axial Conforms The an DSS-ISG-2010-01DSS-ISG-2010-01 3b 3c Rack Model Interfaces Conforms Conforms The rack mode The analysis do JLD-ISG-2015-01, Revision 0: JLD-ISG-2015-01, Compliance with Phase 2 of Order Order EA-13-109, with Licenses Modifying Regard to Reliable Hardened Containment Vents Capable of Operation under Severe Accident Conditions JLD-ISG-12-03, Rev 1: Rev JLD-ISG-12-03, Compliancewith Order EA- 12-051 Concerning Spent Instrumentation Fuel Pool Draft: JLD-ISG-12-04, Performing a Seismic Margin Assessment in Response to Request for the March 2012 Information Letter Draft: JLD-ISG-12-05, Performance of an for Integrated Assessment Flooding Draft: JLD-ISG-12-06, Performing a Tsunami, Surge, or Seiche Hazard Assessment Draft: JLD-ISG-13-01, Estimating Flooding Hazards to Dam Failure due
Tier 2 1.9-203 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 9.1.1 9.1.1 9.1.1 9.1.1 9.1.1 9.1.1 9.1.1 alysis in the code atistically significant increase storageracks. Due to the credit for burnup. 9.1.1 a bias term and an uncertainty bias term a Comments Section are publicly available and are referenced publicly available and are referenced are used in the analysis are either observably does not rely upon cited precedents. 9.1.1 is includes both is includes is includes a trend an is considers the presence of an additional of is considers the presence ysis considers area of applicability in the code of ysis considers area ysis considers fuel handling accidents, rack ysis considers fuel handling accidents, analysis does not take validation. derived from the code validation. validation. in SFP criticalityin analyses. conservativethe presentation or are justified in the of assumption. damage consistent with postulated accidents and full boron dilution.analyses are within All the limits established for normal conditions. assembly alongside the fuel assembly alongside assembliesspacing and the large number of in the base model, there is no st analysis in reactivity. Status AC Title / Description Conformance Table 1.9-4: Conformance with Interim Staff Guidance (ISG) (Continued) (ISG) with Interim Staff Guidance Conformance Table 1.9-4: AC ISG Section/Title DSS-ISG-2010-01DSS-ISG-2010-01 4c 4d Statistical Treatment Fission Products Lumped Conforms Conforms The analys The DSS-ISG-2010-01 4b Trend Analysis Conforms The analys DSS-ISG-2010-01 5b References Conforms Cited references DSS-ISG-2010-01DSS-ISG-2010-01 4e 5a Code-to-Code Comparisons Precedents Conforms Conforms The analysis 9.1.1 DSS-ISG-2010-01 5c Assumptions Conforms Assumptions DSS-ISG-2010-01 4a Applicabilityof Area Conforms The anal DSS-ISG-2010-01 3e Accident Conditions Conforms The anal DSS-ISG-2010-01 3d Conditions Normal Conforms The analys
Tier 2 1.9-204 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 19.0 19.1 19.2 Not Applicable Not Applicable Not Applicable Not Applicable ess reliability core of reliability and ess only to PWRs that are that to PWRs only Comments Section Design certification will addr Design ancontainment heat removal update systems, with reflectrequiredCOL applicant to site-specific by conditions. rule requires an evaluationof proposed PWR This auxiliary feedwater (AFW) systems. The NuScale plant typicalasystem like LWR. AFW design does have an of this ruleNeither theliteral the intent nor language design. the NuScale applies to from large PWRs design differs reactor The NuScale NuScalebecause thedoes not design require or NuScale the Rather, reactor coolant pumps. include primary uses passive natural circulation of the design coolant, eliminating the need for reactor coolant pumps. This guidance is applicable power-operateddesigned with pressurizer relief use power- The NuScale design does not valves. operated relief valves. requirementThis applies only to BWRs.BWRs. Regardless, the requirement applies only to This issue contemplated by this requirement was related to power-operated relief valves.Applicable Not The NuScale design does not use power-operated relief valves. requirementThis applies only to BWRs.Applicable Not Status Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Partially Conforms I Requirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) t pump seal damage t pump seal tion levels (II.K.3.13) ion of the proposed auxiliaryof the proposedion y and risk assessmentto study identify practicable system Regulation Description / Title Conformance feedwater system (II.E.1.1)feedwater system Perform a plant/site-specific probabilistic risk risk probabilistic plant/site-specific a Perform assessment,ofaim the whichis to seek such improvements in the reliability of core and removalcontainment heat systems as are practicaldo not impact significant and and excessively on the plant (II.B.8) Performevaluation an of the potential for and impact of reactor coolan and LOCA (II.K.2.16 following small-break II.K.3.25) probabilityanalysisPerform an of the of a small- LOCA causedbreak by a stuck-open power- operated relief valve (PORV) (II.K.3.2) of providing for separationpressure of highof providing injectioncoolantcore and reactor isolation cooling system initia optimumdetermine the automatic depressurizationdesign modifications system thatneed would eliminate the for manual activation (II.K.3.18) modifications that would reduce challenges and (II.K.3.16) failures of relief valves Table 1.9-5: Conformance with TM Conformance 1.9-5: Table Item 50.34(f)(1)(i) 50.34(f)(1)(iii) 50.34(f)(1)(iv) 50.34(f)(1)(v) Performevaluation an of the safety effectiveness 50.34(f)(1)(vii) a feasibilit Perform 50.34(f)(1)(vi)to a study Perform 50.34(f)(1)(ii)evaluat an Perform
Tier 2 1.9-205 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.1 7.2.13 18.7.2 Not Applicable Not Applicable Not Applicable uded from the information from uded a)(8) and 10a)(8) and CFR 50.34(f), nd development of plant of nd development Comments Section This requirementThis applies only to BWRs.requirementThis applies only to BWRs.Applicable Not requirementThis applies only to BWRs.Applicable Not requirementThis applies only to BWRs.Applicable Not to Pursuant 10 CFR 52.47( is excl paragraph (f)(1)(xii) required to be included in an application for a design certification. Applicable Not Provisions for simulator capability are the responsibilityapplicant of the COL referencing the certified design. The plant procedure improvement program specified (a by this requirement procedures) is the responsibilityof the COL applicant the certifiedreferencing design. safetyThe NuScale display is and indication system control room human-system integrated into the interface design rather than having a separate console. Status Conforms None.Conforms 18.7 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable quirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) (Continued) Issues (NUREG-0933) and Generic 50.34(f)) CFR (10 quirements , valves, accumulators, fect on all core-coolingfect on all e-of-the-art human factor fety parameter display console fety parameter display ation of depressurization Regulation Description / Title Conformance Perform a study of the ef of a study Perform modes under accident conditions of designing andspraycorethe low pressure coolant injection systems (II.K.3.21) for the need to determine a study Perform spaceadditional cooling to ensure reliable long- term operationthe high pressure coolant of injectionand reactor core isolation cooling systems (II.K.3.24) Provide simulator capabilitymodels that correctly to the capability includes room and control the simulate small-break LOCAs(I.A.4.2) Establish a program to improve plant procedures, includewith the program scope to emergency procedures, reliability analyses,factors human management, operator engineering, crisis training,and coordination with INPO and other industry efforts (I.C.9) room Provide, for Commission review, a control designthat reflects stat to committing to fabrication or principles prior ofrevision fabricated control room panels and (I.D.1) layouts (I.D.2) Depressurization System and associated equipment and instrumentation willtheir intended be capable of performing (II.K.3.28) functions control systems methods (II.K.3.45) Table 1.9-5: Conformance with TMI Re Conformance 1.9-5: Table Item 50.34(f)(2)(iv)sa Provide a plant 50.34(f)(2)(i) 50.34(f)(2)(ii) 50.34(f)(2)(iii) 50.34(f)(1)(viii) 50.34(f)(1)(ix) 50.34(f)(1)(x)50.34(f)(1)(xi) the Automatic to ensure that a study Perform 50.34(f)(1)(xii) Provide an evalu evaluation an Perform of alternative hydrogen
Tier 2 1.9-206 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.1 12.4 11.5 12.4 6.3.1 5.2.2 7.2.4 5.4.4 9.3.2 7.2.13 7.2.13 12.3.1 Not Applicable le to the DCA except for except the DCA le to Comments Section R 52.47(a)(8) and 10 CFR 50.34(f), and R 52.47(a)(8) This requirement is applicab This aspects specifyingThe PORV block valve testing. NuScale designnot use power-operated does relief valves. ensure long term core cooling capability. ensure post-accident samplingpost-accident system is not required for 9.3.2 in SRP provided the guidance that provided normal processutilizing the (post- sampling system been satisfied. has accident) Paragraph (f)(2)(ix) is excluded from the information from the is excluded (f)(2)(ix) Paragraph required to be included in an application for a design certification. Status Conforms None. 5.2 Conforms None. 7.1 Conforms None. 12.2 Departure venting of noncondensible The gases is unnecessary to Not ApplicableNot CF to 10 Pursuant Partially Conforms Partially Conforms Partially a C.I.9.3.2, and RG 1.206, 9.3.2, I.6, SRP described by As quirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) (Continued) Issues (NUREG-0933) and Generic 50.34(f)) CFR (10 quirements may, as a result of an may, as a result of that may berequired lief and safety valves lief and re cooling. Systems tore cooling. om the reactor coolant om d shielding design reviews of for hydrogen control that can ty to promptly obtain and to promptly ty Regulation Description / Title Conformance Provide a test programandmodel associatedProvide a test development, and conduct tests to qualify system re reactor coolant block valves (II.D.1) PORV for PWRs, and, position(openorclosed) in the control room (II.D.3) and operableand status of safety systems (I.D.3) noncondensible gases fr system, and other systemssystem, to maintain adequate co achieve this capabilitybeingshall be capable of from the control room and theiroperated operationshallnot lead to an unacceptable loss-of-coolant of probability the in increase accident or an unacceptable challenge to containment integrity. (II.B.1) spaces around systems that containaccident, accident source term radioactive materials, and designnecessary as to access (II.B.2) permit adequate analyze samplescoolantfrom the reactor system containmentthat may containand accident (II.B.3) source term radioactive materials safely accommodate hydrogen generated by the by generated hydrogen safely accommodate fuel-clad metal water equivalent of a 100% reaction (II.B.8) Table 1.9-5: Conformance with TMI Re Conformance 1.9-5: Table Item 50.34(f)(2)(xi) Provide direct indication of relief and safety valve 50.34(f)(2)(v) Provide for automatic indication of the bypassed 50.34(f)(2)(vi) Provide the capability of high point venting of 50.34(f)(2)(vii) radiation an Perform 50.34(f)(2)(viii) Provide capabili 50.34(f)(2)(ix)50.34(f)(2)(x) Provide a system
Tier 2 1.9-207 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.4.5 8.3.1 8.3.2 Not Applicable )(xii) is not considered )(xii) is not aracteristics that vary with with vary that aracteristics LWR. Also, while the DHRS Also, while the DHRS LWR. pplies to the technically pressure, valve position DHRS. Because the language CFR 50.34(f)(2)(xiii) is hot valent to hot standby standby to hot valent to the portion of this of portion to the Comments Section system conditions, which makes DHRS flow a less for the NuScale design. Control useful measurement system parameters other thanroom indication for appropriate to ensure operators flow are more DHRS adequatelymonitorto have the information necessary cooling. These reactor core operation and DHRS DHRS parameters include and reactor coolant system pressure indication, temperature. 10 CFR 50.34(f)(2 applicable to the NuScale and intentof 10 CFR 50.34(f)(1)(ii) donot apply,the requirement is not applicable to the NuScale design. becauseAn exemption would be unnecessary 10 CFR 50.34(f)(1)(ii) only a of the TMI requirements. relevant portions The NuScale design does not have an AFW system ashavedesign does an AFW system not The NuScale at a typical be found would functionsperforms some of theat a of an AFW system DHRS is designed for NuScale- the NuScale PWR, specific transients and system characteristics,its and actuation and indicationis designed accordingly. with regard Specifically requirementflow indication, specifying control room the DHRS operation involves passive natural circulation flow, with flow ch design equi The NuScale ascondition stated in 10 shutdown condition. The NuScale design does not rely maintainon pressurizer heaters to establish and natural circulation in hot shutdown conditions. Status Departure Not Applicable quirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) (Continued) Issues (NUREG-0933) and Generic 50.34(f)) CFR (10 quirements zer heaterpower supply and ticand manual auxiliary Regulation Description / Title Conformance associated motive and controlinterfaces associated motive and power establishsufficientand to maintain natural circulation in hot standby conditions with only onsiteavailable power (II.E.3.1) feedwater (AFW) system initiation, and provide room control in the flow indication system AFW (II.E.1.2) Table 1.9-5: Conformance with TMI Re Conformance 1.9-5: Table Item 50.34(f)(2)(xiii) Provide pressuri 50.34(f)(2)(xii) Provide automa
Tier 2 1.9-208 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 19.2 5.2.5 6.2.4 7.1.5 9.3.6 7.2.13 ns. For 50.34(f)(2)(xiv)(E), accident conditions, with result in generation of of generation in result gh containment pressure Comments Section ve the highest allowable ve (C), and (D). NuScale is requesting an exemption to TMI to TMI an exemption is requesting NuScale and (D). (C), requirement 10 For CFR 50.34(f)(2)(xiv)(E). hi the 50.34(f)(2)(xiv)(D), analytical limit is abo leakcontainment pressure for detection operability. containmentfor initiating point Therefore, the set operation. with normal isolation is compatible Additionally, the containment high pressure analytical limit is subatmospheric, therefore, any pressure setpoint up to and including the analyticalwill limit release to the enviro a prevent the NuScale design differs from that of a traditional vintage design of a TMI-era reactor large water core and resulting because reactor core uncovery, damage, cannot occur without reachinglow the low pressurizer level containment isolation setpoint. The pressurizer is an integral part of the reactor vessel, well above the reactor core, andlocated not reactorconnected to theby core piping. Design basis and hydraulic acceptanceevents meetthermal their 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 resultant core damage, also reliable containment isolation signals, without reliance on isolation An in- on high containment radiation. containment event resulting in core damage or degradation also results in containment isolation on highlow pressurizer level and containment to core damage or leads event that pressure. An degradation also results in containment isolation on low pressurizer level. These features provide a radiological reliable alternative means to prevent the environs. tofrom the CES release Status Departure (B), to 50.34(f)(2)(xiv)(A), conforms design NuScale The quirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) (Continued) Issues (NUREG-0933) and Generic 50.34(f)) CFR (10 quirements Regulation Description / Title Conformance Provide containment isolation systems that (A) ensure all non-essential systems are isolated automatically; (B) ensure each non-essential penetration (except instrument lines) have two in series; (C) do not result in isolation barriers reopening of the containmentvalves isolation on resettingisolation of theuse a signal; (D) containment set point pressure for initiating containment isolation as low as is compatible automatic include (E) with normal operation; and radiationclosing signal on a high all for systems that providethe environs a path to (II.E.4.2) Table 1.9-5: Conformance with TMI Re Conformance 1.9-5: Table Item 50.34(f)(2)(xiv) 50.34(f)(2)(xiv)(A) 50.34(f)(2)(xiv)(B) 50.34(f)(2)(xiv)(C) 50.34(f)(2)(xiv)(D) 50.34(f)(2)(xiv)(E)
Tier 2 1.9-209 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.3 11.5 7.1.1 7.0.4 7.1.2 9.3.2 7.2.13 12.3.4 7.2.13 7.2.13 Not Applicable Not Applicable 2)(xv) is neither required of unique sensitivity thatof unique not needed. In addition, the l PWR ECCS recirculation Comments Section nor included in the NuScale design. This requirement is nor includedNuScale design. This in the design. technically relevantthe NuScale not to The NuScale containment vessel is smaller than a typical containment building, does not contain sub- or require does not not compartments and function as ventingorincorporate system a purge by thiscontemplated requirement. Personnel access during reactor operation is not include pumps, design does and NuScale ECCS not involvedoes a typica mode where ECCS pump performance relies on containment pressure. Thus purge or vent capability as 10prescribed by CFR 50.34(f)( requirement applies only to Babcock and Wilcox This this (B&W) designs. Based on NUREG-0933, applicability was the result B&W reactor designs exhibitedto secondary system transients (both undercooling and overcooling not exhibitdoes such NuScale design events). The sensitivity. Status Conforms None.Conforms None. 6.2.1 4.3.2 Conforms None. 7.1.1 Not Applicable Not Applicable quirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) (Continued) Issues (NUREG-0933) and Generic 50.34(f)) CFR (10 quirements s that provide in the control rumentation adequate for Regulation Description / Title Conformance Capability for containment purging/venting minimizedesigned the purging time to consistent with as low as reasonably achievable (II.E.4.4) (ALARA) design criterion for the allowable Establish of actuation number and cycles of the ECCS reactor protection system with the expected occurrence rates of severe overcooling events (II.E.5.1) and to measure, record, Provide instrumentation readout in the control room: (A) containment pressure, (B) containment(C) water level, containment hydrogen concentration, (D) containment radiationintensity (high level), and nobleat gas effluents(E) all potential, accident for continuous sampling release points. Provide of radioactive iodines and particulates in gaseous effluentspotential from all accident release to analyze and onsite capability for points, and measure these samples (II.F.1) indication of inadequate room an unambiguous core cooling (II.F.2) monitoring plant conditions following an core damage (II.F.3) that includes accident Table 1.9-5: Conformance with TMI Re Conformance 1.9-5: Table Item 50.34(f)(2)(xviii)50.34(f)(2)(xix) Provide instrument Provide inst 50.34(f)(2)(xvi) 50.34(f)(2)(xvii) 50.34(f)(2)(xv)
Tier 2 1.9-210 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.4 11.6 5.4.5 8.1.4 8.3.1 8.3.2 6.3.1 9.3.2 9.3.4 6.4.4 7.2.13 12.3.4 15.0.3 Not Applicable ates its applicability only ef valves and blockef valves valves rd plant design. Aspects of Aspects rd plant design. Comments Section The requirements of 10The requirements of CFR for power 50.34(f)(2)(xx) for pressurizer reli supplies technicallydesign. The relevantare not the NuScale to NuScale design supports an exemption from the portions of 10 CFR 50.34(f)(2)(xx) related to pressurizer level indicators. requirementThis applies only to BWR designs.Applicable Not requirement explicitly st This applicabilitydesigns. This to B&W plant reflectsaspects design that wereidentifiedthe B&W ICS followingof the TMI incident as design/reliability deficiencies, and pertinentdesign. are not the NuScale to requirementThis applies only to BWR designs.Applicable Not requirement is applicableThis to the DCA to extent it is relevantstandato the requirementand thatthis testing are pertinent to operational programs are the responsibility of the COL applicant. Status Conforms None. 11.5 Conforms None. 6.4.1 Departure Not Applicable Not Applicable Not Applicable Not Applicablerequirement This applies only to B&W plant designs. Applicable Not Partially Conforms None. 13.3 Partially Conforms quirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) (Continued) Issues (NUREG-0933) and Generic 50.34(f)) CFR (10 quirements level indicators (II.G.1) level indicators stem (ICS) to include (ICS) stem cording requirements ide containment that ring of in-plantring of radiation and lity to record reactor vessel the reactor protection system, l pathways for radioactivity and pplies for pressurizer relief Technical Support Center and Regulation Description / Title Conformance valves, blockand valves, water level in one location on recorders that meet normal post-accident re (II.K.3.23) Perform a failure modes and effectsof analysis the integrated control sy effects of input and consideration of failures and ICS (II.K.2.9) signals to the output necessary automaticnecessary and manual actions can be takenensure to proper functioning when the operablemain feedwater (II.K.1.22) system is not an anticipatory reactor trip that wouldan anticipatory reactor trip that be actuated on loss of main feedwater and on turbine trip (II.K.2.10) Center (III.A.1.2) Support Operational onsite airborne radioactivity (III.D.3.3) radiation that may lead to control room habitability problemsconditions under accident resulting in an accident source term release (III.D.3.4) designof systems outs contain (or might contain) accident source term radioactive materials (III.D.1.1) Table 1.9-5: Conformance with TMI Re Conformance 1.9-5: Table Item 50.34(f)(2)(xxi)removal Design auxiliary systems such that heat 50.34(f)(2)(xxiii)part of Provide, as 50.34(f)(2)(xx) Provide power su 50.34(f)(2)(xxii) 50.34(f)(2)(xxiv)50.34(f)(2)(xxv) Provide the capabi Provide an onsite 50.34(f)(2)(xxvii) Provide for monito 50.34(f)(2)(xxviii) Evaluate potentia 50.34(f)(2)(xxvi) Provide for leakage control and detection in the
Tier 2 1.9-211 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 6.2 17.4 17.5 19.2 Not Applicable Not Applicable Not Applicable Not Applicable a)(8) and 10a)(8) and CFR 50.34(f), plicable only toapplicants and Comments Section This requirement is the responsibility of the COL of the COL responsibility requirement is the This applicant. requirement is applicableThis to the DCA to extent DCA. of the in support activities to design is relevant it programthis rule specifyingAspects QA of design and analysis, requirementsfor site-specific operational programs, as-built documentation, and constructionand installation responsibility are theof the COL applicant. requirement is not technically relevantThis to the onbased NuScale design.TMI requirement is This traditional largeLWR containment designsthe and need for the requirement, potential,time of as of the to accommodate systems containment venting future vessel containment severe accidents. The NuScale design differscontainment from a typical LWR structure becauseof its high-pressureA 3- capability. footrelative opening to the NuScale containment is discussed in Sectionunnecessary. As the 6.2.1.1.1, peakdesign basis events containment forcalculated remains less than the CNV internal design pressure. As discussed in Sectioncontainment peak 19.2.3, challengepressures do not containment integrity for any analyzed severeaccident progression. (Refer to SectionTR-0716-50424, 2.8). to Pursuant 10 CFR 52.47( is excludedinformation from the(f)(3)(v) paragraph required to be included in an application for a design certification. havedesign does external hydrogen not The NuScale recombiners. holders of reactor facility licenses. Status Conforms None. 3.2 Departure Not Applicable Not Applicable Not Applicablerequirement is ap This Not Applicable Partially Conforms quirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) (Continued) Issues (NUREG-0933) and Generic 50.34(f)) CFR (10 quirements ive procedures for evaluating ive procedures Information - Containment Regulation Description / Title Conformance Ensure that the QA list required by Criterion II in by Criterion QA list required that the Ensure 10Appendix B to CFR includes all SSC 50 (I.F.1) important to safety EstablishQAaspecified Program based on the considerations (I.F.2) recombiners, provide redundant dedicated containment penetrations (II.E.4.1) for design and constructionactivities (II.J.3.1) operating, design, and construction experienceoperating, design, (I.C.5) penetrations, equivalentin size to a single3-foot- diameter opening(II.B.8) Integrity (II.B.8) Integrity Table 1.9-5: Conformance with TMI Re Conformance 1.9-5: Table Item 50.34(f)(3)(iii) 50.34(f)(3)(ii) 50.34(f)(3)(vi)50.34(f)(3)(vii) For plant designs with external hydrogen management Provide a description of the plan 50.34(f)(3)(i) Provide administrat 50.34(f)(3)(iv) Provide one or more dedicated containment 50.34(f)(3)(v) Preliminary Design
Tier 2 1.9-212 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 18.2.3 Not Applicable ned by this guidance is site- boiling water reactors.Applicable Not Comments Section specific. Status Conforms None. 6.3 Not ApplicableIssue is specific to This Not Applicableapplicableis This to currently-operating plants.Not Applicable The information gover Applicable Not quirements (10 CFR 50.34(f)) and Generic Issues (NUREG-0933) (Continued) Issues (NUREG-0933) and Generic 50.34(f)) CFR (10 quirements Regulation Description / Title Conformance Assessment of Debris Accumulation on PWR of Debris Accumulation Assessment Sump Performance Water Reactor Emergency Cooling Water Boiling Concerns System (ECCS) Suction Hazard Estimates in Central and Eastern U.S. on Hazard Estimates in Central and Eastern U.S. Existing Plants Upstream Dam Failures Upstream Dam Table 1.9-5: Conformance with TMI Re Conformance 1.9-5: Table Item Issue 193 Issue 191 Issue 199Updated Probabilistic Seismic Implications of Issue 204 FloodingFollowing of Nuclear Power Plant Sites
Tier 2 1.9-213 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 5.4 8.1 8.2 8.3 Ch 6 9.3.1 8.1.4 8.3.1 8.3.2 8.1.4 8.3.2 8.1.4 7.2.15 Not Applicable Not Applicable er cables that provideer cables that signon does not rely yer packages ensure that Comments Section holders of operating licensesholders for nuclear power reactors. The IAS furnishes both instrument andThe IAS air. IAS service and dr moisture separators the instrument air supplied is dry in accordance with S7.3-R1981. ANS/ISA of the quality standards NuScalePortions relevantto the passive plant design are considered in the design of electrical systems. No safety-related DC systems; however,relevant portions are considered in the design of the non-Class 1E EDSS. Power Plant de NuScale The or risk-significant offsite power for safety-related functions. Gridstudies stability are the responsibility of a COL applicant that referencesNuScale the design certification. NuScale has determined that gas accumulation accident under not impact ECCS will buildup the RCS. It is notconditions. interface with DHRS does connectedthe secondary to system. systems do notsystems do include pow power to equipment with risk-significant or safety-related functions.compliance The scope of GLissues addressed by with the is limited to 2007-01 power cables within the scope of 10 CFR 50.65. Conformance is achieved for cable monitoring by the RGguidance of COL holder applying the 1.218 as discussed in Chapter 8. Status Conforms perience (Generic Letters and Bulletins) Letters (Generic perience Not Applicable Partially Conforms Partially Conforms Partially Conforms Partially Conforms As described in Chapter 8, the electrical power iveness Not Applicableto Applicable elated Logic Circuits Conforms None. 7.2.2 Accumulation in Emergency Core in Emergency Accumulation Underground Power Cable Failures Table 1.9-6: Evaluation of Operating Ex Operating of Evaluation 1.9-6: Table Cooling, Decay Heat Removal, andDecay Containment Heat Removal, Cooling, Spray Systems Resolution of Generic Issue A30, Adequacy Of Adequacy A30, Issue of Generic Resolution to Supplies Pursuant DC Power Safety-Related 10 CFR 50.54(f) Design Processes the Operability of Offsite Power Safety-Related Equipment that Disable Accident Mitigation Systems or Cause Plant Transients. Doc ID Title Conformance GenericLetter 88-15 Generic Letter 91-06 Electric Power Systems - Inadequate Control Over Letter 96-01Generic GenericLetter 2006-02 of Safety-R Testing Grid Reliability and the Impact on Plant Risk 2007-01Bulletin Attent Security Officer Generic Letter 88-14Letter Generic Affecting Problems System Supply Air Instrument Generic Letter 2008-01 Managing Gas Generic Letter 2007-01or Inaccessible
Tier 2 1.9-214 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 8.2.3 Not Applicable 9, which is described9, which in n is demonstrated by the n is demonstrated by 54(hh)(2) is addressedin Comments Section lletin 2011-01 was addressed to existing lletin was addressed to existing Licensees. 2011-01 It required the Licensee to “confirm continue It required The 50.54(hh)(2)…”. with 10 CFR compliance 50. with 10 CFR compliance Section 20.2. Consideration of this bulleti 8- BTP with SRP conformance Section 8.2.3. (Generic Letters and Bulletins) (Continued) Status em Conforms Partially Electric Power Syst Table 1.9-6: Evaluation of Operating Experience Operating of Evaluation 1.9-6: Table Doc ID Title Conformance Bulletin 2011-01 Mitigating Strategies Not Applicable Bu Bulletin 2012-01 Design Vulnerability in
Tier 2 1.9-215 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria ------19.1 19.2 14.3.2 19.2.6 14.3.6 Not Applicable 93-087 93-087 - rs, regulatory guidance, rs, regulatory guidance, rs, regulatory guidance, rs, regulatory guidance, sues (SECYs and Associated and Associated sues (SECYs R 52 and implementing NRC R 52 and implementing NRC R 52 and implementing NRC R 52 and implementing NRC R 52 and implementing NRC Comments Section to evolutionary ALWR designs evolutionary designs ALWR to 084. As a passive ALWR design, the design, passive ALWR a 084. As - 94 - and SECY applicableand is not directly to passive plant designs. Addressed through SECY-90-016 and SECY-93-087. See and SECY-93-087. SECY-90-016 through Addressed Tablefor further information. 1.9-8 This SECY was directed towards evolutionary ALWR issues of certain SECY-90-016 applicability designs. The to passive plants was later established in SECY Incorporated into 10 CF guidance documents. Incorporated into 10 CF guidance documents. Incorporated into 10 CF guidance documents. Incorporated into 10 CF guidance documents. Incorporated into 10 CF guidance documents. Orde Incorporated into NRC and pending rulemaking. Orde Incorporated into NRC and pending rulemaking. Orde Incorporated into NRC and pending rulemaking. Orde Incorporated into NRC and pending rulemaking. NuScale design conforms to the passive plantNuScale design conforms rather and SECY-94-084, guidance of SECY-93-087 See Table than that of SECY-90-016. for further 1.9-8 information. Light Water Reactor Design Is Status Conforms Conforms Conforms Conforms Conforms Conforms Conforms Conforms Conforms Not Applicable pertains SECY-91-078 SRMs) Partially Conforms ght WaterReactor (LWR) ght WaterReactor (LWR) irements Document and Evolutionary Light-Water (LWR) Reactor Evolutionary Light-Water to IssuesCertification and Their Relationship Current Requirements Regulatory under for Design Certification Requirements 10 CFR Part 52 ofPowerResearch the ElectricChapter 11 Institute’s (EPRI's) Requ Li Evolutionary Additional Certification Issues for Design CertificationsITAAC and Combined Licenses Design Review and Requirements for ITAAC of FDA Issuance andResolution of Selected Technical Severe Light-Water Issues for Evolutionary Accident Designs Reactor (LWR) Use of Design Acceptance Criteria During the 10 CFR Certification Reviews Design 52 Part ContainmentThe External Performance Goal, the DefinitionEvents Sequences, of and LWRs Advanced Containment Failure for Advanced Light Water Reactors Water Light Advanced under Part 52 Designs Certified Standard Table 1.9-7: Conformance with Advanced and Evolutionary with Advanced and Evolutionary Conformance Table 1.9-7: Doc ID Title Conformance SECY-89-013SECY-90-016 Design Requirements Related to the Evolutionary SECY-90-241SECY-90-377 SECY-91-074for Design Certification Level of Detail Required SECY-91-078 Designs Prototypefor Advanced Decisions Reactor SECY-91-178 Conforms SECY-91-210 SECY-91-229SECY-91-262 SECY-92-053 Severe AccidentDesign Alternatives for Mitigation SECY-92-092
Tier 2 1.9-216 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria - 5.4 8.1 8.2 8.3 8.4 8.1 8.2 8.3 8.4 19.3 19.3 9.2.5 15.0.4 Appendix 9A sues (SECYs and Associated and Associated sues (SECYs R 52 and implementing NRC R 52 and implementing NRC R 52 and implementing NRC n any RTNSS equipment. equipment. n any RTNSS Comments Section Incorporated into 10 CF The NuScale Fire Protection guidance documents. System does not contai However, Section C, Safe Shutdown Requirements, of the SECY discusses the stable shutdown conditionfor passive ALWR which is applicablethe NuScale to Power Plant. Incorporated into 10 CF guidance documents. Incorporated into 10 CF guidance documents. Light Water Reactor Design Is Status Conforms Conforms Not Applicable Site-specific requirements. Applicable Not Not Applicable None. 13.3 See Table 1.9-8. None. - Partially Conforms SRMs) (Continued) ionary and Passive Light- ionary and Framework for Nuclear Power Framework for ed Technical and Licensinged Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Designs Reactor (ALWR) PolicyTechnical and Issues Associatedthe with of Non-Safety Systems in Regulatory Treatment Passive Plant Design (RTNSS) Proposed Options to Address Lessons-Learned Review of the U.S. Nuclear Regulatory Commissions Force-On-ForceProgram Inspection - Staff Requirements Memorandum to in Response COMGEA/COMWCO-14-0001 Regulatory Treatment of Non-Safety Systems in Regulatory Treatment Passive Plant Designs Issues Relating to Evolut Relating to Issues Designs Water-Reactor Plant Emergency Preparedness Oversight Table 1.9-7: Conformance with Advanced and Evolutionary with Advanced and Evolutionary Conformance Table 1.9-7: Doc ID Title Conformance SECY-93-087 SECY-94-084 SECY-94-302SECY-95-132 Source-Term-Relat SECY-14-038 PolicyTechnical and Issues Associated with SECY-14-088 Performance-Based
Tier 2 1.9-217 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 15.8 19.2 19.2.2 Not Applicable Not Applicable 016 - 90 - Comments Section design relies on diversity within the module protection system within the associated the risk reduce to (MPS) events.with ATWS loops use external Design does not down conditionand no drain for refueling. SECY relevance of theThe SBO issue to passive ALWR designs was deferred to and addressedin and SECY- SECY-94-084 of Section F The NuScale designconforms 95-132. theseguidanceto the passive plant documents. Status Conforms None. Appendix 9A Conforms None. 15.0.3 Conforms None. 9.3.4 ConformsConforms None.Conforms None. None.Conforms None. 19.2 19.2.3 19.1 19.2.4 Not Applicable Not Applicable Partially ConformsNuScale The Partially Conforms 6.2.5 cal, and Licensing Issues Pertaining to Evolutionary and g area and quenching g area and Advanced Light-Water Reactor Designs" Reactor Light-Water Advanced : Position on the current practices : Position on the current s to mitigate the effects of a loss of all features necessary to ensure a high necessary features ensure a high degree of protection from ensure a high Incorporationjudgment of engineering tion on acceptable design features to to features design tion on acceptable ion: Position for a dedicated vent Position for a dedicated ion: able requirements to measure and mitigate y criteria for coolin y obability of an ISLOCA. obability system and othermanagement schemes ensure to safeshutdownthe reactor. of and a more realistic source term in designthat deviates from thesiting requirements in 10 CFR 100. Use of a Physically-Based Source Term: Physically-Based Source a Use of (ATWS) SCRAM without Transient Anticipated achieveagainstand design featuresan a high degree of protection to ATWS. Mid-Loop Operation: Position on design reliabilityRHR systems in PWR. degree of of Station BlackoutPosition on method (SBO): Intersystem LOCA: Position on acceptable design practices and preventative measures to minimize the pr AC power. fuel zirconium with reaction water a to due produced hydrogen of effects the cladding. ability regarding corium interaction with concrete. regarding ability prevent the event of a high-pressure core melt ejection. high-pressurea core melt the event of prevent failureprobabilities or otheranalyses to containment. the penetration to precludecontainmentaresultingover- containment failure from pressurization event. Table 1.9-8: Conformance with SECY-93-087, "Policy, Techni "Policy, with SECY-93-087, Conformance Table 1.9-8: Issue Description Conformance I.D I.A I.EProtection: Positions Fire on design configuration and featuresfire protection the I.F I.B I.C I.G Hydrogen Control: Position on accept I.HDebris Coolability: Acceptabilit Core I.II.J Ejection: Posi Core Melt High-Pressure I.KPerformance: Position on acceptable Containment containment conditional Containment Vent Penetrat Dedicated
Tier 2 1.9-218 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria all 3.7 3.3 19.1 19.2 3.6.3 Not Applicable Comments Section By settingBy the OBE toof 1/3 the SSE it from the design decoupled is evaluation process. endorsed use the latest NuScale codes and standards or others on basis. case by case The NuScale electrical system design to the passiveconforms plant Section G. guidance of SECY-94-084, linesand FW MS to theLBB is applied inside containment. The FSAR uses the maximum tornado wind230 speed of mph found in RGRevision 1.76 rather than the 1 outdated 300 mph guidance found in SECY-93-087. Status Conforms None. 3.9.6 Conforms Conforms Conforms None.Conforms None. 19.1.5 15.0.3 Conforms None. 19.2.3 Not Applicable Not Applicable Applicable to BWRs. Not Applicable Partially Conforms Partially Conforms None. 6.2.6 cal, and Licensing Issues Pertaining to Evolutionary and tor Designs" (Continued) Designs" tor developed or modified ters: Positionof on use proposed ide: (1) an alternate power source to stem to channel fission product releases Position on periodic testing to confirm on periodic testing to Position on use of recently irements related to plant features oneconnected offsite directly circuit to Advanced Light-Water Reac Light-Water Advanced ves and performance of seismicPRA. BWRs: Position on the staffs defined defined staffs BWRs: Position on the inally addressed by SECY-91-078 that SECY-91-078 inally addressed by on the applicability of environmental on the applicability of Earthquake: Position on the applicability of the Position on the applicability of Earthquake: operability of safety-related pumps and valves. approachseismic forand classification of the main steam line in both evolutionary passive BWRs. for a design basis tornado. Elimination of Operating-Basis design and the possibilityin OBE the designin of decouplingof the OBE and SSE safety systems. Industry Codes and Standards: Position Electricalorig Distribution: Positions prov that an evolutionary ALWR specified nonsafety-related loads,and (2) at least each redundant safetydivision withnointervening nonsafety-related buses. Design Parame Curves and Hazard Seismic generic bounding seismic hazard cur Classification of Main Steam Lines in ContainmentBypass: Position design against containment bypass. on ALWR the containment sy Specifically, failure of through the suppressionpool,of or the failure passive containment cooling heat exchanger tubescontainment.outsidewater in large pools of design codes and industry standardsALWR designs in that have not been for acceptabilityreviewed by the NRC. testing (prior to rule change). qualification and quality assurance requ provided only for severe-accident protection. Table 1.9-8: Conformance with SECY-93-087, "Policy, Techni "Policy, with SECY-93-087, Conformance Table 1.9-8: Issue Description Conformance II.C I.N Testing of In-Service Pumps and Valves: II.E II.G II.A II.B I.L Survivability: Equipment Position II.F to be used speed tornado wind maximum on the Position Design Basis: Tornado II.D Leak-Before-Break: Position on use of leak-before-break concept. Conforms II.H Leak Containment Rate Testing: Position on testing duration for Type C leak rate I.M
Tier 2 1.9-219 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 19.1 9.3.2 All FSAR Sections Comments Section 1.206, C.I.9.3.2, a post-accident C.I.9.3.2, 1.206, sampling systemnot is required that the guidance providedprovided for utilizingin SRP 9.3.2 the normal process sampling system (post- beenaccident) has satisfied. Status Conforms and RG I.6, by SRP 9.3.2, As described Conforms None. Conforms None.ConformsConforms None. None.Conforms 14.3 None.Conforms 17.4 None. 19.1 ConformsConforms None. None. 7.1.5 15.6 19.1 7.2.13 Conforms None. 19.2.6 Not Applicable Information Only. Applicable Not cal, and Licensing Issues Pertaining to Evolutionary and tor Designs" (Continued) Designs" tor esign certification iring that analysis of multiple SG Tube multiple SG that analysis of iring ity: Position on recommending that ity: Position on recommending Events: Position onthe inclusion of on requiring conversion of the design on requiring conversion of basis that needs to be addressed as a design reliability assurance program. reliability a design sign approval/designsign an certification of the application for design certification of application the in-depth and diversityof instrumentation nal design approval/d rnatives (SAMDAs): Position on the Position on rnatives (SAMDAs): Certification: No guidance provided; guidance No Certification: rtification submittal with depth of detail depth of rtification submittal with demonstrate that a nuclear power plant Advanced Light-Water Reac Light-Water Advanced m systems are necessary to minimize the and operatesconsistent the with design ed; information only Conforms None. 1.5 e design certification review. e design a plant-specific PRA dissolvedaccordancein hydrogen, oxygen, and chlorideapplicable with regulations. Level of Detail: Position on a design ce Level of similar to that in an FSAR. ITAAC: Position on providing ITAAC to certified design is built areferencing certification. of External and Analyses PRAs Site-Specific external event analysis beyond the design th during PRA the plant of part Accident Mitigation Design Severe Alte fi of the as part consideration of SAMDA AgainstDefenseControl Common-Mode Failures Digital Instrumentation and in ofPosition on the use defense- Systems: Failures: Position on requ Tube Multiple SG in tubes be included SG 2 to 5 Failures of passive ALWRs. Beyond Position DesignPRA Certification: certification PRA into Control Room Annunciator (Alarm) Reliabil scope, objectives, and implementation of scope, de final the of part as systems control and advanced reactor. additional requirements for ALWR alar problems experienced by operating nuclear power plants of an advanced reactor. advanced an of information only. Table 1.9-8: Conformance with SECY-93-087, "Policy, Techni "Policy, with SECY-93-087, Conformance Table 1.9-8: Issue Description Conformance II.J II.KII.L provid No guidance Prototyping: II.I Post-AccidentSampling System:Position on the requiredcapability analyze to II.MII.N ReliabilityAssurance Program: Position on providing a descriptionpurpose, of II.O II.R II.S II.T II.Q II.PRulemaking Related to Design Generic
Tier 2 1.9-220 Revision 2 NuScale Final Safety Analysis Report Conformance with Regulatory Criteria 7.1 3.1.4 5.4.3 18.10 15.0.3 Comments Section BWR requirement.using the non-safety related drain system containment flood and allow the containment to to flood cooldown to cold conditions for disconnectionand transferNPMs. of Not Applicable andDuring shutdown NPM and decay heatmovement, residual removal is provided by heat convection and conduction from the the reactor pool via reactor to RCS, flooded containment, and the RPV and containment vessel walls. Status Conforms None. 15.0.3 Conforms None.Conforms None.Conforms The provisions of this SECY are met by 19.3 15.0.0 Conforms None.Conforms None.Conforms None. 6.5.3 13.3 18.7 cal, and Licensing Issues Pertaining to Evolutionary and tor Designs" (Continued) Designs" tor ves to be evaluated ation to be performed andbeation toperformed that a fully Systems in Passive Designs: Position Position Designs: Passive in Systems k valves) to cause val k valves) to cause wn since non-safety RHR systems do not in determining the acceptability criteriaacceptabilityin determining the the staff redefining some passive failures redefining the staff ion on containment spray systems for ion on containment spray systems ning: Position on simplifying off-site simplifying Position on ning: on usingon non-safety grade active cooling appropriate analytical methods (i.e., dose appropriate analytical Advanced Light-Water Reac Light-Water Advanced rdance with regulatory standards. y of the SBWRy of Applicable Not systems to bring a reactor to cold shutdo cold to reactor a bring to systems comply with the guidance of 1.139 or branch technical position 5-1 passive ALWRs. integratedfunctional control room prototype is necessary passive forplant control room designs to demonstrate that functionstasks and are integrated properly into theman/machineinterface decisions. Regulatory Treatment of Active Nonsafety of components as active failures (i.e., chec in previouslicensingthanmuch more stringent manner review in a Position Safe Shutdown Requirements: Radionuclide Attenuation: Position on fission product removal processes inside pipingbuilding by and the secondary effects and holdupby natural containment systems in addition to commission posit passiveemergencyofdesigns due to the estimated low probability planning of core damage of such designs. the PassivePlant Control Room Operator:Role of on Commission position sufficient man-in-the-loop testing and evalu on the proposed staff approach for resolving the regulatory treatment of the activenon-safety systems in passive ALWRs. limits and accident duration) to be used controlfor room habitably in acco Table 1.9-8: Conformance with SECY-93-087, "Policy, Techni "Policy, with SECY-93-087, Conformance Table 1.9-8: Issue Description Conformance III.EIII.F on Position Habitability: Room Control III.A III.BIII.C on of Passive Failure: Position Definition III.D Stabilit Thermal-Hydraulic III.GIII.H Plan Emergency of Offsite Simplification
Tier 2 1.9-221 Revision 2 NuScale Final Safety Analysis Report Nuclear Power Plants to be Operated on Multi-Unit Sites
1.10 Nuclear Power Plants to be Operated on Multi-Unit Sites
COL 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.
Tier 2 1.10-1 Revision 2