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mPower\TM\ Design-Specific Review Standard

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Start Preamble

AGENCY:

Nuclear Regulatory Commission.

ACTION:

Design-Specific Review Standard (DSRS) for the mPowerTM Design; request for comment.

SUMMARY:

The U.S. Nuclear Regulatory Commission (NRC) is soliciting public comment on the Design-Specific Review Standard (DSRS) for the mPowerTM design (mPowerTM DSRS). The purpose of the mPowerTM DSRS is to more fully integrate the use of risk insights into the review of a design certification (DC), an early site permit (ESP) or a combined license (COL) that incorporates the mPowerTM design.

DATES:

Submit comments by August 16, 2013. Comments received after this date will be considered, if it is practical to do so, but the Commission is able to ensure consideration only for comments received on or before this date.

ADDRESSES:

You may submit comment 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 http://www.regulations.gov and search for Docket ID NRC-2013-0089. Address questions about NRC dockets to Carol Gallagher; telephone: 301-492-3668; email: Carol.Gallagher@nrc.gov. For technical questions, contact the individual(s) 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: TWB-05-B01M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
  • Fax comments to: RADB at 301-492-3446.Start Printed Page 28259

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

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FOR FURTHER INFORMATION CONTACT:

Ms. Yanely Malave, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-1519 or email at Yanely.Malave@nrc.gov.

End Further Info End Preamble Start Supplemental Information

SUPPLEMENTARY INFORMATION:

I. Accessing Information and Submitting Comments

A. Accessing Information

Please refer to Docket ID NRC-2013-0089 when contacting the NRC about the availability of information regarding this document. You may access information related to this document, which the NRC possesses and is publicly-available, by the following methods:

  • Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0089.
  • NRC's Agencywide Documents Access and Management System (ADAMS): You may access publicly-available documents online in the NRC Library at http://www.nrc.gov/​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 pdr.resource@nrc.gov . The ADAMS accession number for each document referenced in this notice (if that document is available in ADAMS) is provided the first time that a document is referenced and also in the table included in this notice. The mPowerTM DSRS Scope and Safety Review Matrix is available in ADAMS under Accession No. ML13088A252.
  • 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-2013-0089 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

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 posts all comment submissions at http://www.regulations.gov as well as entering 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 submissions into ADAMS.

II. Further Information

A. Background

In 2010, the Commission provided direction to the staff on the preparation for, and review of, small modular reactor (SMR) applications, with a near-term focus on integral pressurized water reactor (iPWR) designs. The Commission directed the 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 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 Design-Specific Review Standard. This DSRS for the mPowerTM design is the result of the implementation of the Commission's direction.

B. Design-Specific Review Standard (DSRS) for the mPowerTM Design

As part of the mPowerTM Design-Specific Review Plan, the Office of New Reactors has issued the mPowerTM Design-Specific Review Standard Scope and Safety Review Matrix to reflect the integration of risk insights into the review of applications submitted for the mPowerTM DC, and ESPs or COLs that incorporate the mPowerTM design under 10 CFR Part 52. The mPowerTM DSRS reflects current staff review methods and practices based on the integration of risk insights and, where appropriate, lessons learned from NRC reviews of DC and COL applications completed since the last revision of the Standard Review Plan.

The NRC staff is issuing this notice to solicit public comment on the mPowerTM DSRS Scope and Safety Review Matrix (Matrix), and the individual mPowerTM-specific DSRS sections. Specifically, we request comment on the sufficiency of the proposed mPowerTM review scope encompassed by the Matrix, and comment on technical content of the individual mPowerTM DSRS sections identified in the table below that were revised or developed to incorporate design-specific review guidance based on features of the mPowerTM reactor design. We are not, however, soliciting detailed technical comments on NUREG-0800 Standard Review Plan sections that are designated with the applicability “A) Use SRP Section as-is . . .” in the Matrix unless their adequacy for review of the mPowerTM design is in question.

SectionDesign-specific review standard titleADAMS No.
MatrixmPowerTM DSRS Scope and Safety Review MatrixML13088A252
2.4.0Hydrology ReviewML12355A691
2.4.1Hydrologic DescriptionML12221A023
2.4.2FloodsML12221A024
2.4.3Probable Maximum Flood (PMF) on Streams and RiversML12221A025
2.4.4Potential Dam FailuresML12221A026
2.4.5Probable Maximum Surge and Seiche FloodingML12221A027
2.4.6Probable Maximum Tsunami FloodingML12221A028
2.4.7Ice EffectsML12221A017
2.4.9Channel DiversionsML12221A018
2.4.10Flooding Protection RequirementsML12221A019
2.4.12GroundwaterML12221A020
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2.4.13Accidental Releases of Radioactive Liquid Effluents in Ground and Surface WatersML12221A021
2.4.14Technical Specifications and Emergency Operation RequirementsML12221A022
3.2.1Seismic ClassificationML12272A013
3.2.2System Quality Group ClassificationML12272A015
3.3.1Severe Wind LoadingML12324A156
3.3.2Extreme Wind Loads (Tornado and Hurricane Loads)ML12324A166
3.4.1Internal Flood Protection for Onsite Equipment FailureML12312A148
3.4.2Protection of Structures Against Flood From External SourcesML12324A190
3.5.1.1Internally Generated Missiles (Outside Containment)ML12313A158
3.5.1.2Internally Generated Missiles (Inside Containment)ML12313A396
3.5.1.3Turbine MissilesML12272A209
3.5.1.4Missiles Generated by Extreme WindsML12313A399
3.5.1.5Site Proximity Missiles (Except Aircraft)ML12318A151
3.5.1.6Aircraft HazardsML12318A198
3.5.2Structures, Systems, and Components To Be Protected From Externally Generated MissilesML12313A457
3.5.3Barrier Design ProceduresML12222A003
3.6.2Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of PipingML12230A013
3.7.1Seismic Design ParametersML13099A204
3.7.2Seismic System AnalysisML13099A205
3.7.3Seismic Subsystem AnalysisML13099A209
3.8.2Steel ContainmentML13099A298
3.8.3Concrete and Steel Internal Structures of Steel ContainmentsML13099A312
3.8.4Other Seismic Category I StructuresML13099A316
3.8.5FoundationsML13099A319
3.9.1Special Topics for Mechanical ComponentsML12272A018
3.9.4Control Rod Drive SystemsML12272A020
3.9.5Reactor Pressure Vessel InternalsML12272A077
3.9.6Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic RestraintsML12272A217
3.11Environmental Qualification of Mechanical and Electrical EquipmentML12277A018
3.13Threaded Fasteners—ASME Code Class 1, 2, and 3ML12272A214
BTP 3-4Postulated Rupture Locations in Fluid System Piping Inside and Outside ContainmentML12272A102
4.2Fuel System DesignML12235A168
4.3Nuclear DesignML12353A188
4.4Thermal and Hydraulic DesignML12319A580
4.5.1Control Rod Drive Structural MaterialsML12326A740
4.5.2Reactor Internal and Core Support Structure MaterialsML12272A006
4.6Functional Design of Control Rod Drive SystemML12353A182
5.2.1.1Compliance With the Codes and Standards Rule, 10 CFR 50.55aML12272A091
5.2.1.2Applicable Code CasesML12272A096
5.2.3Reactor Coolant Pressure Boundary MaterialsML12272A007
5.2.5Reactor Coolant Pressure Boundary Leakage DetectionML12313A468
5.3.1Reactor Vessel MaterialsML12272A008
5.3.2Pressure-Temperature Limits, Upper-shelf Energy, and Pressurized Thermal ShockML12272A009
5.3.3Reactor Vessel IntegrityML12272A010
5.4.2.1Steam Generator MaterialsML12272A244
5.4.2.2Steam Generator ProgramML12272A245
5.4.7Residual Heat Removal (RHR) SystemML12319A582
BTP 5-4Design Requirements of the Residual Heat Removal SystemML12275A020
6.1.1Engineered Safety Features MaterialsML12276A107
6.1.2Protective Coating Systems (Paints)—Organic MaterialsML12272A246
6.2.1Containment Functional DesignML12276A117
6.2.1.1mPower iPWR ContainmentML12227A377
6.2.1.2Subcompartment AnalysisML12230A014
6.2.1.3Mass and Energy Release Analysis for Postulated Loss of Coolant AccidentsML12230A034
6.2.1.4Mass and Energy Release Analysis for Postulated Secondary System Pipe RupturesML12230A037
6.2.2Containment Heat Removal SystemsML12276A118
6.2.4Containment Isolation SystemML12276A120
6.2.5Combustible Gas Control in ContainmentML12276A124
6.2.6Containment Leakage TestingML12276A127
6.2.7Fracture Prevention of Containment Pressure BoundaryML12278A103
6.4Control Room Habitability SystemML12272A225
6.6Inservice Inspection and Testing of Class 2 and 3 ComponentsML12284A064
BTP 6-1PH for Emergency Coolant Water for Pressurized Water ReactorsML12222A198
BTP 6-2Minimum Containment Pressure Model for PWR ECCS Performance EvaluationML12227A380
BTP 6-4Containment Purging During Normal Plant OperationsML12227A384
7.0 (DSRS)Instrumentation and Controls—Introduction and Overview of Review ProcessML12314A197
7.1 (DSRS)Instrumentation and Controls—Fundamental Design PrinciplesML12313A479
7.2 (DSRS)Instrumentation and Controls—System CharacteristicsML12314A201
7.0 APP A (DSRS)Instrumentation and Controls—Hazard AnalysisML12318A200
7.0 APP B (DSRS)Instrumentation and Controls—System ArchitectureML12318A201
7.0 APP C (DSRS)Instrumentation and Controls—SimplicityML12318A204
7.0 APP D (DSRS)Instrumentation and Controls—ReferencesML12318A205
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8.1Electric Power/IntroductionML12269A005
8.2Offsite Power SystemML12269A006
8.3.1A C Power Systems (Onsite)ML12269A010
8.3.2D C Power Systems (Onsite)ML12269A011
8.4Station BlackoutML12269A015
BTP 8-2Use of Diesel-Generator Sets for PeakingML12269A016
BTP 8-3Stability of Offsite Power SystemsML12269A017
BTP 8-6Adequacy of Station Electric Distribution System VoltagesML12269A018
9.1.3Spent Fuel Pool Cooling and Cleanup SystemML12319A063
9.2.1Station Service Water SystemML12319A068
9.2.2Reactor Auxiliary Cooling Water SystemsML12325A088
9.2.4Potable and Sanitary Water SystemsML12319A091
9.2.5Ultimate Heat SinkML12319A423
9.2.6Condensate Storage FacilitiesML12270A276
9.3.2Process and Post Accident Sampling SystemsML12170A005
9.3.3Equipment and Floor Drainage SystemML12319A437
9.4.1Control Room Area Ventilation SystemML12276A130
9.4.2Spent Fuel Pool Area Ventilation SystemML12272A229
9.4.3Reactor Service Building HVAC SystemsML12276A133
9.4.4Turbine Area Ventilation SystemML12221A117
9.5.2Communications SystemsML12277A361
9.5.3Lighting SystemsML12319A516
10.2Turbine GeneratorML12320A111
10.2.3Turbine Rotor IntegrityML12272A247
10.3Main Steam Supply SystemML12320A134
10.3.6Steam and Feedwater System MaterialsML12272A004
10.4.1Main CondensersML12320A139
10.4.2Main Condenser Evacuation SystemML12320A146
10.4.3Turbine Gland Sealing SystemML12320A157
10.4.4Turbine Bypass SystemML12320A161
10.4.5Circulating Water SystemML12320A172
10.4.6Condensate Cleanup SystemML12272A242
10.4.7Condensate and Feedwater SystemML12320A183
11.1Source TermsML12222A292
11.2Liquid Waste Management SystemsML12257A228
11.3Gaseous Waste Management SystemsML12257A227
11.4Solid Waste Management SystemsML12257A223
11.5Process and Effluent Radiological Monitoring Instrumentation and Sampling SystemsML12258A115
11.6Guidance on Instrumentation and Control Design Features for Process and Effluent Radiological Monitoring, and Area Radiation and Airborne Radioactivity MonitoringML13023A089
BTP 11-3Design Guidance for Solid Radioactive Waste Management Systems Installed in Light-Water -Cooled Nuclear Power Reactor PlantsML12222A293
BTP 11-5Postulated Radioactive Releases Due to a Waste Gas System Leak or FailureML12222A294
12.1Assuring that Occupational Radiation Exposures Are As Low As Is Reasonably AchievableML12222A295
12.2Radiation SourcesML12222A296
12.3—12.4Radiation Protection Design FeaturesML12269A175
12.5Operational Radiation Protection ProgramML12257A224
14.2Initial Plant Test Program—Design Certification and New License ApplicantsML12121A037
14.3.2Structural and Systems Engineering—Inspections, Tests, Analyses, and Acceptance CriteriaML12272A243
14.3.4Reactor Systems—Inspections, Tests, Analyses, and Acceptance CriteriaML12353A174
14.3.5Instrumentation and Controls—Inspections, Tests, Analyses, and Acceptance CriteriaML12325A091
14.3.6Electrical Systems—Inspections, Tests, Analyses, and Acceptance CriteriaML12320A188
14.3.7Plant Systems—Inspections, Tests, Analyses, and Acceptance CriteriaML12320A195
14.3.8Radiation Protection—Inspections, Tests, Analyses, and Acceptance CriteriaML12257A225
15.0Introduction—Transient and Accident AnalysesML12275A026
15.0.2Review of Transient and Accident Analysis MethodsML12207A098
15.0.3Design Basis Accident Radiological Consequence Analyses for Advanced Light Water ReactorsML12257A226
15.1.5Steam System Piping Failures Inside and Outside of ContainmentML12207A108
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)ML12319A584
15.2.6Loss of Nonemergency AC Power to the Station AuxiliariesML12319A587
15.2.7Loss of Normal Feedwater FlowML12250A248
15.2.8Feedwater System Pipe Breaks Inside and Outside Containment (PWR)ML12319A668
15.3.1-15.3.2Loss of Forced Reactor Coolant Flow Including Trip of Pump Motor and Flow Controller MalfunctionsML12319A585
15.3.3-15.3.4Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft BreakML12319A586
15.4.1Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup ConditionML12240A005
15.4.2Uncontrolled Control Rod Assembly Withdrawal at PowerML12242A102
15.4.10Startup of an Inactive Pump or Pumps at an Incorrect Temperature, and Flow Controller Malfunction causing an Increase in Core Flow RateML12261A399
15.5.1-15.5.2Inadvertent Operation of ECCS and Reactor Coolant Inventory and Purification System (RCI) Malfunction that Increases Reactor Coolant InventoryML12319A575
15.6.1Inadvertent Opening of a Pressurizer Safety Valve, or an Automatic Depressurization ValveML12250A318
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15.6.5Loss of Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure BoundaryML12319A576
15.8Anticipated Transients Without ScramML12319A577
15.9.AThermal Hydraulic StabilityML12261A042
16.0Technical SpecificationsML12270A277
Start Signature

Dated at Rockville, Maryland, this 6th day of May, 2013.

For the Nuclear Regulatory Commission.

Yanely Malave,

Project Manager Small Modular Reactor Licensing Branch 1, Division of Advanced Reactors and Rulemaking, Office of New Reactors.

End Signature End Supplemental Information

[FR Doc. 2013-11394 Filed 5-13-13; 8:45 am]

BILLING CODE 7590-01-P