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NuScale Power, LLC, Design-Specific Review Standard and Safety Review Matrix

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Nuclear Regulatory Commission.


Design-specific review standard; request for comment.


The U.S. Nuclear Regulatory Commission (NRC) is soliciting public comment on the Design-Specific Review Standard (DSRS) and Safety Review Matrix for the NuScale Power, LLC, design (NuScale DSRS Scope and Safety Review Matrix). The purpose of the NuScale DSRS is to provide guidance to NRC staff in performing safety reviews where existing NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition,” Standard Review Plans (SRP) have been modified by the staff specifically for the NuScale design, or do not address unique features of the NuScale design. The DSRS also allows NRC staff to more fully integrate the use of design-specific risk insights into the review of the NuScale design certification application (DC) or an early site permit (ESP) or combined license (COL) application that references the NuScale design.


Submit comments by August 31, 2015. Comments received after this date will be considered, if it is practical to do so, but the NRC is able to ensure consideration only for comments received on or before this date.


You may submit comments by any of the following methods (unless this document describes a different method for submitting comments on a specific subject):

  • Federal Rulemaking Web site: Go to and search for Docket ID NRC-2015-0160. Address questions about NRC dockets to Carol Gallagher; telephone: 301-415-3463; email: For technical questions, contact the individual listed in the FOR FURTHER INFORMATION CONTACT section of this document.
  • Mail comments to: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration, Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

For additional direction on obtaining information and submitting comments, see “Obtaining Information and Submitting Comments” in the SUPPLEMENTARY INFORMATION section of this document.

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Jenny Gallo, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-7367; email:

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I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2015-0160 when contacting the NRC about the availability of information regarding this document. You may obtain publicly-available information related to this action by any of the following methods:

  • Federal Rulemaking Web site: Go to and search for Docket ID NRC-2015-0160.
  • NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly-available documents online in the ADAMS Public Documents collection at​reading-rm/​adams.html. To begin the search, select “ADAMS Public Documents” and then select “Begin Web-based ADAMS Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to The ADAMS accession number for each document referenced (if it is available in ADAMS) is provided the first time that it is mentioned in the SUPPLEMENTARY INFORMATION section. The NuScale DSRS Scope and Safety Review Matrix is available in ADAMS under Accession No. ML15156B063.
  • NRC's PDR: You may examine and purchase copies of public documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

Please include Docket ID NRC-2015-0160 in your comment submission.

The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at as well as enter the comment submissions into ADAMS. The NRC does not routinely edit comment submissions to remove identifying or contact information.

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information that they do not want to be publicly disclosed in their comment submission. Your request should state that the NRC does not routinely edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment into ADAMS.

II. Further Information

A. Background

In the Staff Requirements Memorandum (SRM) COMGBJ-10-0004/COMGEA-10-0001, “Use of Risk Insights to Enhance the Safety Focus of Small Modular Reactor Reviews,” dated August 31, 2010 (ADAMS Accession No. ML102510405), the Commission provided direction to the NRC staff on the preparation for, and review of, small modular reactor (SMR) applications, with a near-term focus on integral pressurized-water reactor designs. The Commission directed the NRC staff to more fully integrate the use of risk insights into pre-application activities and the review of applications and, consistent with regulatory requirements and Commission policy statements, to align the review focus and resources to risk-significant structures, systems, and components and other aspects of the design that contribute most to safety in order to enhance the effectiveness and efficiency of the review process. The Commission directed the NRC staff to develop a design-specific, risk-informed review plan for each SMR design to address pre-application and application review activities. An important part of this review plan is the DSRS. The DSRS for the NuScale design is the result of the implementation of the Commission's direction.

B. DSRS for the NuScale Design

The NuScale DSRS reflects current NRC staff safety review methods and practices which integrate risk insights and, where appropriate, lessons learned from the NRC's reviews of DC and COL applications completed since the last revision of the NUREG-0800, SRP Introduction, Part 2, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: Light-Water Small Modular Reactor Edition,” January 2014 (ADAMS Accession No. ML13207A315). The NuScale DSRS Scope and Safety Matrix provides a complete list of SRP sections and identifies which SRP sections will be used for DC, COL, or ESP reviews concerning the NuScale design; which SRP sections are not applicable to the Start Printed Page 37313NuScale design; and which new DSRS sections are design-specific to NuScale. The NuScale DSRS Scope and Safety Review Matrix is available in ADAMS under Accession No. ML15156B063.

The NRC staff is soliciting public comment on the NuScale DSRS Scope and Safety Review Matrix and the individual NuScale-specific DSRS sections referenced in the table below. Specifically, the NRC requests comment on the sufficiency of the scope of the proposed NuScale review, as encompassed by the Safety Review Matrix, and on the technical content of the individual NuScale-specific DSRS sections identified in the table below. These sections were revised from the relative SRP sections or developed to incorporate design-specific review guidance based on features of the NuScale design. The NRC is not soliciting general comments on NUREG-0800 sections that are designated with the applicability “A) Use SRP Section” in the Safety Review Matrix, but specific comments on the adequacy of these NUREG-0800 sections for use in the review of the NuScale design certification application will be considered.

SectionDesign-specific review standard titleADAMS Accession No.
MatrixNuScale Power, LLC DSRS Scope and Safety Review MatrixML15156B063
3.11Environmental Qualification of Mechanical and Electrical EquipmentML15131A247
3.13Threaded Fasteners—ASME Code Class 1, 2, and 3ML15084A277
3.3.1Offsite Power SystemML15071A259
3.3.2Tornado LoadsML15071A267
3.4.1Internal Flood Protection for Onsite Equipment FailuresML15139A112
3.4.2Analysis ProceduresML15071A324 Generated Missiles (Outside Containment)ML15139A081 Generated Missiles (Inside Containment)ML15139A096 MissilesML15070A248 Generated by Tornadoes and Extreme WindsML15139A121
3.5.2Structures, Systems, and Components to be Protected from Externally-Generated MissilesML15139A102
3.5.3Barrier Design ProceduresML15071A273
3.7.1Seismic Design ParametersML15084A279
3.7.2Seismic System AnalysisML15084A177
3.7.3Seismic Subsystem AnalysisML15131A340
3.8.2Steel ContainmentML15131A373
3.8.4Other Seismic Category I StructuresML15118A151
4.2Fuel System DesignML15132A517
4.3Nuclear DesignML15125A374
4.4Thermal and Hydraulic DesignML15131A427
4.5.2Reactor Internal and Core Support Structure MaterialsML15070A325
4.6Functional Design of Control Rod Drive SystemML15119A111
5.2.2Overpressure ProtectionML15118A931
5.2.4Reactor Coolant Pressure Boundary Inservice Inspection and TestingML15125A305
5.2.5Reactor Coolant Pressure Boundary Leakage DetectionML15132A194
5.3.1Reactor Vessel MaterialsML15070A457
5.3.2Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal ShockML15070A468
5.3.3Reactor Vessel IntegrityML15070A462
5.4Rx Coolant System Component and Subsystem DesignML15126A156 Generator MaterialsML15131A376 Generator ProgramML15070A562
5.4.7Residual Heat Removal (RHR) SystemML15131A360
5-4 BTPDesign Requirements of the RHR SystemML15132A524
6.1.1Engineered Safety Features MaterialsML15070A567
6.1.2Protective Coating Systems (Paints)—Organic MaterialsML15071A372
6-1 BTPpH for Emergency Coolant Water for PWRsML15125A369
6.2.1Containment Functional DesignML15118A922 Dry Containments, Including Sub-atmospheric ContainmentsML15118A264 and Energy Release Analysis for Postulated Loss-of-Coolant Accidents (LOCAs)ML15112A134 and Energy Release Analysis for Postulated Secondary System Pipe RupturesML15118A293
6.2.2Containment Heat Removal SystemsML15131A341
6.2.4Containment Isolation SystemML15119A087
6.2.5Combustible Gas Control in ContainmentML15119A090
6.2.6Containment Leakage TestingML15119A084
6.2.7Fracture Prevention of Containment Pressure BoundaryML15112A517
6.3Emergency Core Cooling SystemML15125A322
6.6Inservice Inspection and Testing of Class 2 and 3 ComponentsML15127A136
7.0Instrumentation and Controls—Introduction and Overview of Review ProcessML15125A340
7.0, AInstrumentation and Controls—Hazard AnalysisML15132A583
7.0, BInstrumentation and Controls—System ArchitectureML15132A603
7.0, CInstrumentation and Controls—SimplicityML15132A611
7.0, DInstrumentation and Controls—ReferencesML15132A618
7.1I&C—Fundamental Design PrinciplesML15125A335
7.2Instrumentation and Controls—System CharacteristicsML15125A360
8.1Electric Power—IntroductionML15146A269
8.2Offsite Power SystemML15125A425
8-2 BTPUse of Diesel-Generator Sets for PeakingMl15131A386
8.3.1AC Power Systems (Onsite)ML15125A384
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8.3.2DC Power Systems (Onsite)ML15125A386
8-3 BTPStability of Offsite Power SystemsML15125A390
8.4Station BlackoutML15126A149
8-6 BTPAdequacy of Station Electric Distribution System VoltagesML15131A461
9.1.2New and Spent Fuel StorageML15125A307
9.1.3Spent Fuel Pool Cooling and Cleanup SystemML15146A034
9.2.6Condensate Storage FacilitiesML15131A245
9.3.2Process and Post-Accident Sampling SystemsML15131A298
9.3.4Chemical and Volume Control System (PWR) (Including Boron Recovery System)ML15131A305
9.3.6Containment Evacuation and Flooding SystemsML15112A190
9.5.2Communications SystemsML15084A403
9.5.3Lighting SystemsML15112A148
10.2Turbine GeneratorML15126A086
10.2.3Turbine Rotor IntegrityML15127A046
10.3Main Steam Supply SystemML15131A329
10.4.1Main CondensersML15127A049
10.4.2Main Condenser Evacuation SystemML15127A349
10.4.3Turbine Gland Sealing SystemML15126A477
10.4.4Turbine Bypass SystemML15131A417
10.4.5Circulating Water SystemML15126A467
10.4.6Condensate Cleanup SystemML15118A943
10.4.7Condensate and Feedwater SystemML15126A470
10.4.10Auxiliary Boiler SystemML15131A261
11.1Source TermsML15112A526
11.2Liquid Waste Management SystemML15124A607
11.3Gaseous Waste Management SystemML15112A694
11.4Solid Waste Management SystemML15119A057
11.5Process and Effluent Radiological Monitoring Instrumentation and Sampling SystemsML15118A609
11.6Guidance on I&C Design Features for Process and Effluent Radiological Monitoring and Area Radiation and Airborne Radioactivity MonitoringML15125A367
12.2Radiation SourcesML15070A194
12.3-12.4Radiation Protection Design FeaturesML15070A204
12.5Operational Radiation Protection ProgramML15070A210
14.2Initial Plant Test Program—Design Certification and New License ApplicantsML15084A407
14.3.2Structural and Systems Engineering—Inspections, Tests, Analyses, and Acceptance CriteriaML15084A411
14.3.4Reactor Systems—Inspections, Tests, Analyses, and Acceptance CriteriaML15125A294
14.3.5Instrumentation and Controls—Inspections, Tests, Analyses, and Acceptance CriteriaML15127A383
14.3.6Electrical Systems—Inspections, Tests, Analyses, and Acceptance CriteriaML15127A373
14.3.7Plant Systems—Inspections, Tests, Analyses, and Acceptance CriteriaML15131A328
15.0Introduction—Transient and Accident AnalysesML15125A297
15.0.3Design Basis Accidents Radiological Consequence Analyses for Advanced Light Water ReactorsML15127A387
15.1.1-15.1.4Decrease in FW Temperature, Increase in FW Flow, Increase in Steam Flow, and Inadvertent Opening of a Steam Generator Relief or Safety ValveML15127A391
15.1.5Steam System Piping Failures Inside and Outside of Containment (PWR)ML15125A317
15.1.6Loss of Containment VacuumML15127A395
15.2.1-15.2.5Loss of External Load; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve (BWR); and Steam Pressure Regulator Failure (Closed)ML15127A400
15.2.6Loss of Non-Emergency AC Power to the Station AuxiliariesML15125A292
15.2.7Loss of Normal Feedwater FlowML15125A293
15.2.8Feedwater System Pipe Breaks Inside and Outside Containment (PWR)ML15118A927
15.4.1Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup ConditionML15118A482
15.4.2Uncontrolled Control Rod Assembly Withdrawal at PowerML15118A600
15.4.3Control Rod Misoperation (System Malfunction or Operator Error)ML15131A364
15.4.6Inadvertent Decrease in Boron Concentration in the Reactor Coolant (PWR)ML15118A474
15.5.1-15.5.2Chemical and Volume Control System Malfunction that Increases Reactor Coolant InventoryML15125A463
15.6.5LOCAs Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure BoundaryML15131A334
15.6.6Inadvertent Opening of a PWR Pressurizer Pressure Relief ValveML15125A467
15.9AThermal-hydraulic StabilityML15131A311
16.0Technical SpecificationsML15131A316
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Dated at Rockville, Maryland, this 23rd day of June 2015.

For the Nuclear Regulatory Commission.

Jenny M. Gallo,

Project Manager, Small Modular Reactor Licensing Branch, Division of Advanced Reactors and Rulemaking, Office of New Reactors.

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[FR Doc. 2015-16034 Filed 6-29-15; 8:45 am]