78 FR 93 pgs. 28258-28262 - mPowerTMDesign-Specific Review Standard
Type: NOTICEVolume: 78Number: 93Pages: 28258 - 28262
Docket number: [NRC-2013-0089]
FR document: [FR Doc. 2013-11394 Filed 5-13-13; 8:45 am]
Agency: Nuclear Regulatory Commission
Official PDF Version: PDF Version
NUCLEAR REGULATORY COMMISSION
[NRC-2013-0089]
mPowerTM Design-Specific Review Standard
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.
For additional direction on accessing information and submitting comments, see "Accessing Information and Submitting Comments" in the SUPPLEMENTARY INFORMATION section of this document.
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.
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.
| Section | Design-specific review standard title | ADAMS No. |
|---|---|---|
| Matrix | mPowerTM DSRS Scope and Safety Review Matrix | ML13088A252 |
| 2.4.0 | Hydrology Review | ML12355A691 |
| 2.4.1 | Hydrologic Description | ML12221A023 |
| 2.4.2 | Floods | ML12221A024 |
| 2.4.3 | Probable Maximum Flood (PMF) on Streams and Rivers | ML12221A025 |
| 2.4.4 | Potential Dam Failures | ML12221A026 |
| 2.4.5 | Probable Maximum Surge and Seiche Flooding | ML12221A027 |
| 2.4.6 | Probable Maximum Tsunami Flooding | ML12221A028 |
| 2.4.7 | Ice Effects | ML12221A017 |
| 2.4.9 | Channel Diversions | ML12221A018 |
| 2.4.10 | Flooding Protection Requirements | ML12221A019 |
| 2.4.12 | Groundwater | ML12221A020 |
| 2.4.13 | Accidental Releases of Radioactive Liquid Effluents in Ground and Surface Waters | ML12221A021 |
| 2.4.14 | Technical Specifications and Emergency Operation Requirements | ML12221A022 |
| 3.2.1 | Seismic Classification | ML12272A013 |
| 3.2.2 | System Quality Group Classification | ML12272A015 |
| 3.3.1 | Severe Wind Loading | ML12324A156 |
| 3.3.2 | Extreme Wind Loads (Tornado and Hurricane Loads) | ML12324A166 |
| 3.4.1 | Internal Flood Protection for Onsite Equipment Failure | ML12312A148 |
| 3.4.2 | Protection of Structures Against Flood From External Sources | ML12324A190 |
| 3.5.1.1 | Internally Generated Missiles (Outside Containment) | ML12313A158 |
| 3.5.1.2 | Internally Generated Missiles (Inside Containment) | ML12313A396 |
| 3.5.1.3 | Turbine Missiles | ML12272A209 |
| 3.5.1.4 | Missiles Generated by Extreme Winds | ML12313A399 |
| 3.5.1.5 | Site Proximity Missiles (Except Aircraft) | ML12318A151 |
| 3.5.1.6 | Aircraft Hazards | ML12318A198 |
| 3.5.2 | Structures, Systems, and Components To Be Protected From Externally Generated Missiles | ML12313A457 |
| 3.5.3 | Barrier Design Procedures | ML12222A003 |
| 3.6.2 | Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping | ML12230A013 |
| 3.7.1 | Seismic Design Parameters | ML13099A204 |
| 3.7.2 | Seismic System Analysis | ML13099A205 |
| 3.7.3 | Seismic Subsystem Analysis | ML13099A209 |
| 3.8.2 | Steel Containment | ML13099A298 |
| 3.8.3 | Concrete and Steel Internal Structures of Steel Containments | ML13099A312 |
| 3.8.4 | Other Seismic Category I Structures | ML13099A316 |
| 3.8.5 | Foundations | ML13099A319 |
| 3.9.1 | Special Topics for Mechanical Components | ML12272A018 |
| 3.9.4 | Control Rod Drive Systems | ML12272A020 |
| 3.9.5 | Reactor Pressure Vessel Internals | ML12272A077 |
| 3.9.6 | Functional Design, Qualification, and Inservice Testing Programs for Pumps, Valves, and Dynamic Restraints | ML12272A217 |
| 3.11 | Environmental Qualification of Mechanical and Electrical Equipment | ML12277A018 |
| 3.13 | Threaded Fasteners-ASME Code Class 1, 2, and 3 | ML12272A214 |
| BTP 3-4 | Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment | ML12272A102 |
| 4.2 | Fuel System Design | ML12235A168 |
| 4.3 | Nuclear Design | ML12353A188 |
| 4.4 | Thermal and Hydraulic Design | ML12319A580 |
| 4.5.1 | Control Rod Drive Structural Materials | ML12326A740 |
| 4.5.2 | Reactor Internal and Core Support Structure Materials | ML12272A006 |
| 4.6 | Functional Design of Control Rod Drive System | ML12353A182 |
| 5.2.1.1 | Compliance With the Codes and Standards Rule, 10 CFR 50.55a | ML12272A091 |
| 5.2.1.2 | Applicable Code Cases | ML12272A096 |
| 5.2.3 | Reactor Coolant Pressure Boundary Materials | ML12272A007 |
| 5.2.5 | Reactor Coolant Pressure Boundary Leakage Detection | ML12313A468 |
| 5.3.1 | Reactor Vessel Materials | ML12272A008 |
| 5.3.2 | Pressure-Temperature Limits, Upper-shelf Energy, and Pressurized Thermal Shock | ML12272A009 |
| 5.3.3 | Reactor Vessel Integrity | ML12272A010 |
| 5.4.2.1 | Steam Generator Materials | ML12272A244 |
| 5.4.2.2 | Steam Generator Program | ML12272A245 |
| 5.4.7 | Residual Heat Removal (RHR) System | ML12319A582 |
| BTP 5-4 | Design Requirements of the Residual Heat Removal System | ML12275A020 |
| 6.1.1 | Engineered Safety Features Materials | ML12276A107 |
| 6.1.2 | Protective Coating Systems (Paints)-Organic Materials | ML12272A246 |
| 6.2.1 | Containment Functional Design | ML12276A117 |
| 6.2.1.1 | mPower iPWR Containment | ML12227A377 |
| 6.2.1.2 | Subcompartment Analysis | ML12230A014 |
| 6.2.1.3 | Mass and Energy Release Analysis for Postulated Loss of Coolant Accidents | ML12230A034 |
| 6.2.1.4 | Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures | ML12230A037 |
| 6.2.2 | Containment Heat Removal Systems | ML12276A118 |
| 6.2.4 | Containment Isolation System | ML12276A120 |
| 6.2.5 | Combustible Gas Control in Containment | ML12276A124 |
| 6.2.6 | Containment Leakage Testing | ML12276A127 |
| 6.2.7 | Fracture Prevention of Containment Pressure Boundary | ML12278A103 |
| 6.4 | Control Room Habitability System | ML12272A225 |
| 6.6 | Inservice Inspection and Testing of Class 2 and 3 Components | ML12284A064 |
| BTP 6-1 | PH for Emergency Coolant Water for Pressurized Water Reactors | ML12222A198 |
| BTP 6-2 | Minimum Containment Pressure Model for PWR ECCS Performance Evaluation | ML12227A380 |
| BTP 6-4 | Containment Purging During Normal Plant Operations | ML12227A384 |
| 7.0 (DSRS) | Instrumentation and Controls-Introduction and Overview of Review Process | ML12314A197 |
| 7.1 (DSRS) | Instrumentation and Controls-Fundamental Design Principles | ML12313A479 |
| 7.2 (DSRS) | Instrumentation and Controls-System Characteristics | ML12314A201 |
| 7.0 APP A (DSRS) | Instrumentation and Controls-Hazard Analysis | ML12318A200 |
| 7.0 APP B (DSRS) | Instrumentation and Controls-System Architecture | ML12318A201 |
| 7.0 APP C (DSRS) | Instrumentation and Controls-Simplicity | ML12318A204 |
| 7.0 APP D (DSRS) | Instrumentation and Controls-References | ML12318A205 |
| 8.1 | Electric Power/Introduction | ML12269A005 |
| 8.2 | Offsite Power System | ML12269A006 |
| 8.3.1 | A C Power Systems (Onsite) | ML12269A010 |
| 8.3.2 | D C Power Systems (Onsite) | ML12269A011 |
| 8.4 | Station Blackout | ML12269A015 |
| BTP 8-2 | Use of Diesel-Generator Sets for Peaking | ML12269A016 |
| BTP 8-3 | Stability of Offsite Power Systems | ML12269A017 |
| BTP 8-6 | Adequacy of Station Electric Distribution System Voltages | ML12269A018 |
| 9.1.3 | Spent Fuel Pool Cooling and Cleanup System | ML12319A063 |
| 9.2.1 | Station Service Water System | ML12319A068 |
| 9.2.2 | Reactor Auxiliary Cooling Water Systems | ML12325A088 |
| 9.2.4 | Potable and Sanitary Water Systems | ML12319A091 |
| 9.2.5 | Ultimate Heat Sink | ML12319A423 |
| 9.2.6 | Condensate Storage Facilities | ML12270A276 |
| 9.3.2 | Process and Post Accident Sampling Systems | ML12170A005 |
| 9.3.3 | Equipment and Floor Drainage System | ML12319A437 |
| 9.4.1 | Control Room Area Ventilation System | ML12276A130 |
| 9.4.2 | Spent Fuel Pool Area Ventilation System | ML12272A229 |
| 9.4.3 | Reactor Service Building HVAC Systems | ML12276A133 |
| 9.4.4 | Turbine Area Ventilation System | ML12221A117 |
| 9.5.2 | Communications Systems | ML12277A361 |
| 9.5.3 | Lighting Systems | ML12319A516 |
| 10.2 | Turbine Generator | ML12320A111 |
| 10.2.3 | Turbine Rotor Integrity | ML12272A247 |
| 10.3 | Main Steam Supply System | ML12320A134 |
| 10.3.6 | Steam and Feedwater System Materials | ML12272A004 |
| 10.4.1 | Main Condensers | ML12320A139 |
| 10.4.2 | Main Condenser Evacuation System | ML12320A146 |
| 10.4.3 | Turbine Gland Sealing System | ML12320A157 |
| 10.4.4 | Turbine Bypass System | ML12320A161 |
| 10.4.5 | Circulating Water System | ML12320A172 |
| 10.4.6 | Condensate Cleanup System | ML12272A242 |
| 10.4.7 | Condensate and Feedwater System | ML12320A183 |
| 11.1 | Source Terms | ML12222A292 |
| 11.2 | Liquid Waste Management Systems | ML12257A228 |
| 11.3 | Gaseous Waste Management Systems | ML12257A227 |
| 11.4 | Solid Waste Management Systems | ML12257A223 |
| 11.5 | Process and Effluent Radiological Monitoring Instrumentation and Sampling Systems | ML12258A115 |
| 11.6 | Guidance on Instrumentation and Control Design Features for Process and Effluent Radiological Monitoring, and Area Radiation and Airborne Radioactivity Monitoring | ML13023A089 |
| BTP 11-3 | Design Guidance for Solid Radioactive Waste Management Systems Installed in Light-Water -Cooled Nuclear Power Reactor Plants | ML12222A293 |
| BTP 11-5 | Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure | ML12222A294 |
| 12.1 | Assuring that Occupational Radiation Exposures Are As Low As Is Reasonably Achievable | ML12222A295 |
| 12.2 | Radiation Sources | ML12222A296 |
| 12.3-12.4 | Radiation Protection Design Features | ML12269A175 |
| 12.5 | Operational Radiation Protection Program | ML12257A224 |
| 14.2 | Initial Plant Test Program-Design Certification and New License Applicants | ML12121A037 |
| 14.3.2 | Structural and Systems Engineering-Inspections, Tests, Analyses, and Acceptance Criteria | ML12272A243 |
| 14.3.4 | Reactor Systems-Inspections, Tests, Analyses, and Acceptance Criteria | ML12353A174 |
| 14.3.5 | Instrumentation and Controls-Inspections, Tests, Analyses, and Acceptance Criteria | ML12325A091 |
| 14.3.6 | Electrical Systems-Inspections, Tests, Analyses, and Acceptance Criteria | ML12320A188 |
| 14.3.7 | Plant Systems-Inspections, Tests, Analyses, and Acceptance Criteria | ML12320A195 |
| 14.3.8 | Radiation Protection-Inspections, Tests, Analyses, and Acceptance Criteria | ML12257A225 |
| 15.0 | Introduction-Transient and Accident Analyses | ML12275A026 |
| 15.0.2 | Review of Transient and Accident Analysis Methods | ML12207A098 |
| 15.0.3 | Design Basis Accident Radiological Consequence Analyses for Advanced Light Water Reactors | ML12257A226 |
| 15.1.5 | Steam System Piping Failures Inside and Outside of Containment | ML12207A108 |
| 15.2.1-15.2.5 | Loss of External Load; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve (BWR); and Steam Pressure Regulator Failure (Closed) | ML12319A584 |
| 15.2.6 | Loss of Nonemergency AC Power to the Station Auxiliaries | ML12319A587 |
| 15.2.7 | Loss of Normal Feedwater Flow | ML12250A248 |
| 15.2.8 | Feedwater System Pipe Breaks Inside and Outside Containment (PWR) | ML12319A668 |
| 15.3.1-15.3.2 | Loss of Forced Reactor Coolant Flow Including Trip of Pump Motor and Flow Controller Malfunctions | ML12319A585 |
| 15.3.3-15.3.4 | Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break | ML12319A586 |
| 15.4.1 | Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition | ML12240A005 |
| 15.4.2 | Uncontrolled Control Rod Assembly Withdrawal at Power | ML12242A102 |
| 15.4.10 | Startup of an Inactive Pump or Pumps at an Incorrect Temperature, and Flow Controller Malfunction causing an Increase in Core Flow Rate | ML12261A399 |
| 15.5.1-15.5.2 | Inadvertent Operation of ECCS and Reactor Coolant Inventory and Purification System (RCI) Malfunction that Increases Reactor Coolant Inventory | ML12319A575 |
| 15.6.1 | Inadvertent Opening of a Pressurizer Safety Valve, or an Automatic Depressurization Valve | ML12250A318 |
| 15.6.5 | Loss of Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary | ML12319A576 |
| 15.8 | Anticipated Transients Without Scram | ML12319A577 |
| 15.9.A | Thermal Hydraulic Stability | ML12261A042 |
| 16.0 | Technical Specifications | ML12270A277 |
Dated at Rockville, Maryland, this 6th day of May, 2013.
For the Nuclear Regulatory Commission.
Yanely Malave,
Project ManagerSmall Modular Reactor Licensing Branch 1,Division of Advanced Reactors and Rulemaking,Office of New Reactors.
[FR Doc. 2013-11394 Filed 5-13-13; 8:45 am]
BILLING CODE 7590-01-P