NuScale Power, LLC, Design-Specific Review Standard and Safety Review Matrix, 37312-37314 [2015-16034]

Download as PDF 37312 Federal Register / Vol. 80, No. 125 / Tuesday, June 30, 2015 / Notices NUCLEAR REGULATORY COMMISSION [NRC–2015–0160] NuScale Power, LLC, Design-Specific Review Standard and Safety Review Matrix Nuclear Regulatory Commission. ACTION: Design-specific review standard; request for comment. AGENCY: 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. DATES: 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. ADDRESSES: 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 https://www.regulations.gov and search for Docket ID NRC–2015–0160. Address questions about NRC dockets to Carol Gallagher; telephone: 301–415–3463; email: Carol.Gallagher@nrc.gov. 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 asabaliauskas on DSK5VPTVN1PROD with NOTICES SUMMARY: VerDate Sep<11>2014 17:34 Jun 29, 2015 Submitting Comments’’ in the section of this document. FOR FURTHER INFORMATION CONTACT: Jenny Gallo, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001; telephone: 301–415–7367; email: NuScale-DSRS@ nrc.gov. SUPPLEMENTARY INFORMATION: 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. I. Obtaining Information and Submitting Comments II. Further Information SUPPLEMENTARY INFORMATION Jkt 235001 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 https://www.regulations.gov and search for Docket ID NRC–2015–0160. • NRC’s Agencywide Documents Access and Management System (ADAMS): You may obtain publiclyavailable documents online in the ADAMS Public Documents collection at https://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 (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 https:// www.regulations.gov 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 PO 00000 Frm 00094 Fmt 4703 Sfmt 4703 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: LightWater 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 E:\FR\FM\30JNN1.SGM 30JNN1 Federal Register / Vol. 80, No. 125 / Tuesday, June 30, 2015 / Notices NuScale 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. ADAMS Accession No. asabaliauskas on DSK5VPTVN1PROD with NOTICES Section Design-specific review standard title Matrix ................. 3.11 .................... 3.13 .................... 3.3.1 ................... 3.3.2 ................... 3.4.1 ................... 3.4.2 ................... 3.5.1.1 ................ 3.5.1.2 ................ 3.5.1.3 ................ 3.5.1.4 ................ 3.5.2 ................... 3.5.3 ................... 3.7.1 ................... 3.7.2 ................... 3.7.3 ................... 3.8.2 ................... 3.8.4 ................... 3.8.5 ................... 4.2 ...................... 4.3 ...................... 4.4 ...................... 4.5.2 ................... 4.6 ...................... 5.2.2 ................... 5.2.4 ................... 5.2.5 ................... 5.3.1 ................... 5.3.2 ................... 5.3.3 ................... 5.4 ...................... 5.4.2.1 ................ 5.4.2.2 ................ 5.4.7 ................... 5–4 BTP ............ 6.1.1 ................... 6.1.2 ................... 6–1 BTP ............ 6.2.1 ................... 6.2.1.1.A ............ 6.2.1.3 ................ 6.2.1.4 ................ 6.2.2 ................... 6.2.4 ................... 6.2.5 ................... 6.2.6 ................... 6.2.7 ................... 6.3 ...................... 6.6 ...................... 7.0 ...................... 7.0, A ................. 7.0, B ................. 7.0, C ................. 7.0, D ................. 7.1 ...................... 7.2 ...................... 8.1 ...................... 8.2 ...................... 8–2 BTP ............ 8.3.1 ................... NuScale Power, LLC DSRS Scope and Safety Review Matrix ....................................................................... Environmental Qualification of Mechanical and Electrical Equipment ............................................................. Threaded Fasteners—ASME Code Class 1, 2, and 3 ..................................................................................... Offsite Power System ....................................................................................................................................... Tornado Loads ................................................................................................................................................. Internal Flood Protection for Onsite Equipment Failures ................................................................................. Analysis Procedures ......................................................................................................................................... Internally Generated Missiles (Outside Containment) ..................................................................................... Internally Generated Missiles (Inside Containment) ........................................................................................ Turbine Missiles ................................................................................................................................................ Missiles Generated by Tornadoes and Extreme Winds .................................................................................. Structures, Systems, and Components to be Protected from Externally-Generated Missiles ........................ Barrier Design Procedures ............................................................................................................................... Seismic Design Parameters ............................................................................................................................. Seismic System Analysis ................................................................................................................................. Seismic Subsystem Analysis ............................................................................................................................ Steel Containment ............................................................................................................................................ Other Seismic Category I Structures ............................................................................................................... Foundations ...................................................................................................................................................... Fuel System Design ......................................................................................................................................... Nuclear Design ................................................................................................................................................. Thermal and Hydraulic Design ......................................................................................................................... Reactor Internal and Core Support Structure Materials .................................................................................. Functional Design of Control Rod Drive System ............................................................................................. Overpressure Protection ................................................................................................................................... Reactor Coolant Pressure Boundary Inservice Inspection and Testing .......................................................... Reactor Coolant Pressure Boundary Leakage Detection ................................................................................ Reactor Vessel Materials ................................................................................................................................. Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock ................................... Reactor Vessel Integrity ................................................................................................................................... Rx Coolant System Component and Subsystem Design ................................................................................ Steam Generator Materials .............................................................................................................................. Steam Generator Program ............................................................................................................................... Residual Heat Removal (RHR) System ........................................................................................................... Design Requirements of the RHR System ...................................................................................................... Engineered Safety Features Materials ............................................................................................................. Protective Coating Systems (Paints)—Organic Materials ................................................................................ pH for Emergency Coolant Water for PWRs ................................................................................................... Containment Functional Design ....................................................................................................................... PWR Dry Containments, Including Sub-atmospheric Containments ............................................................... Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents (LOCAs) .............................. Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures ................................. Containment Heat Removal Systems .............................................................................................................. Containment Isolation System .......................................................................................................................... Combustible Gas Control in Containment ........................................................................................................ Containment Leakage Testing ......................................................................................................................... Fracture Prevention of Containment Pressure Boundary ................................................................................ Emergency Core Cooling System .................................................................................................................... Inservice Inspection and Testing of Class 2 and 3 Components .................................................................... Instrumentation and Controls—Introduction and Overview of Review Process .............................................. Instrumentation and Controls—Hazard Analysis ............................................................................................. Instrumentation and Controls—System Architecture ....................................................................................... Instrumentation and Controls—Simplicity ........................................................................................................ Instrumentation and Controls—References ..................................................................................................... I&C—Fundamental Design Principles .............................................................................................................. Instrumentation and Controls—System Characteristics ................................................................................... Electric Power—Introduction ............................................................................................................................ Offsite Power System ....................................................................................................................................... Use of Diesel-Generator Sets for Peaking ....................................................................................................... AC Power Systems (Onsite) ............................................................................................................................ VerDate Sep<11>2014 17:34 Jun 29, 2015 Jkt 235001 PO 00000 Frm 00095 Fmt 4703 Sfmt 4703 37313 E:\FR\FM\30JNN1.SGM 30JNN1 ML15156B063 ML15131A247 ML15084A277 ML15071A259 ML15071A267 ML15139A112 ML15071A324 ML15139A081 ML15139A096 ML15070A248 ML15139A121 ML15139A102 ML15071A273 ML15084A279 ML15084A177 ML15131A340 ML15131A373 ML15118A151 ML15132A186 ML15132A517 ML15125A374 ML15131A427 ML15070A325 ML15119A111 ML15118A931 ML15125A305 ML15132A194 ML15070A457 ML15070A468 ML15070A462 ML15126A156 ML15131A376 ML15070A562 ML15131A360 ML15132A524 ML15070A567 ML15071A372 ML15125A369 ML15118A922 ML15118A264 ML15112A134 ML15118A293 ML15131A341 ML15119A087 ML15119A090 ML15119A084 ML15112A517 ML15125A322 ML15127A136 ML15125A340 ML15132A583 ML15132A603 ML15132A611 ML15132A618 ML15125A335 ML15125A360 ML15146A269 ML15125A425 Ml15131A386 ML15125A384 37314 Federal Register / Vol. 80, No. 125 / Tuesday, June 30, 2015 / Notices ADAMS Accession No. Section Design-specific review standard title 8.3.2 ................... 8–3 BTP ............ 8.4 ...................... 8–6 BTP ............ 9.1.2 ................... 9.1.3 ................... 9.2.6 ................... 9.3.2 ................... 9.3.4 ................... 9.3.6 ................... 9.5.2 ................... 9.5.3 ................... 10.2 .................... 10.2.3 ................. 10.3 .................... 10.4.1 ................. 10.4.2 ................. 10.4.3 ................. 10.4.4 ................. 10.4.5 ................. 10.4.6 ................. 10.4.7 ................. 10.4.10 ............... 11.1 .................... 11.2 .................... 11.3 .................... 11.4 .................... 11.5 .................... 11.6 .................... DC Power Systems (Onsite) ............................................................................................................................ Stability of Offsite Power Systems ................................................................................................................... Station Blackout ................................................................................................................................................ Adequacy of Station Electric Distribution System Voltages ............................................................................. New and Spent Fuel Storage ........................................................................................................................... Spent Fuel Pool Cooling and Cleanup System ............................................................................................... Condensate Storage Facilities ......................................................................................................................... Process and Post-Accident Sampling Systems ............................................................................................... Chemical and Volume Control System (PWR) (Including Boron Recovery System) ...................................... Containment Evacuation and Flooding Systems ............................................................................................. Communications Systems ................................................................................................................................ Lighting Systems .............................................................................................................................................. Turbine Generator ............................................................................................................................................ Turbine Rotor Integrity ...................................................................................................................................... Main Steam Supply System ............................................................................................................................. Main Condensers ............................................................................................................................................. Main Condenser Evacuation System ............................................................................................................... Turbine Gland Sealing System ........................................................................................................................ Turbine Bypass System ................................................................................................................................... Circulating Water System ................................................................................................................................. Condensate Cleanup System ........................................................................................................................... Condensate and Feedwater System ................................................................................................................ Auxiliary Boiler System ..................................................................................................................................... Source Terms ................................................................................................................................................... Liquid Waste Management System ................................................................................................................. Gaseous Waste Management System ............................................................................................................. Solid Waste Management System ................................................................................................................... Process and Effluent Radiological Monitoring Instrumentation and Sampling Systems ................................. Guidance on I&C Design Features for Process and Effluent Radiological Monitoring and Area Radiation and Airborne Radioactivity Monitoring. Radiation Sources ............................................................................................................................................ Radiation Protection Design Features ............................................................................................................. Operational Radiation Protection Program ...................................................................................................... Initial Plant Test Program—Design Certification and New License Applicants ............................................... Structural and Systems Engineering—Inspections, Tests, Analyses, and Acceptance Criteria ..................... Reactor Systems—Inspections, Tests, Analyses, and Acceptance Criteria .................................................... Instrumentation and Controls—Inspections, Tests, Analyses, and Acceptance Criteria ................................. Electrical Systems—Inspections, Tests, Analyses, and Acceptance Criteria .................................................. Plant Systems—Inspections, Tests, Analyses, and Acceptance Criteria ........................................................ Introduction—Transient and Accident Analyses .............................................................................................. Design Basis Accidents Radiological Consequence Analyses for Advanced Light Water Reactors .............. Decrease in FW Temperature, Increase in FW Flow, Increase in Steam Flow, and Inadvertent Opening of a Steam Generator Relief or Safety Valve. Steam System Piping Failures Inside and Outside of Containment (PWR) .................................................... Loss of Containment Vacuum .......................................................................................................................... Loss of External Load; Turbine Trip; Loss of Condenser Vacuum; Closure of Main Steam Isolation Valve (BWR); and Steam Pressure Regulator Failure (Closed). Loss of Non-Emergency AC Power to the Station Auxiliaries ......................................................................... Loss of Normal Feedwater Flow ...................................................................................................................... Feedwater System Pipe Breaks Inside and Outside Containment (PWR) ...................................................... Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition ........... Uncontrolled Control Rod Assembly Withdrawal at Power .............................................................................. Control Rod Misoperation (System Malfunction or Operator Error) ................................................................. Inadvertent Decrease in Boron Concentration in the Reactor Coolant (PWR) ............................................... Chemical and Volume Control System Malfunction that Increases Reactor Coolant Inventory ..................... LOCAs Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary. Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve ................................................................ Thermal-hydraulic Stability ............................................................................................................................... Technical Specifications ................................................................................................................................... 12.2 .................... 12.3–12.4 ........... 12.5 .................... 14.2 .................... 14.3.2 ................. 14.3.4 ................. 14.3.5 ................. 14.3.6 ................. 14.3.7 ................. 15.0 .................... 15.0.3 ................. 15.1.1–15.1.4 ..... 15.1.5 ................. 15.1.6 ................. 15.2.1–15.2.5 ..... 15.2.6 ................. 15.2.7 ................. 15.2.8 ................. 15.4.1 ................. 15.4.2 ................. 15.4.3 ................. 15.4.6 ................. 15.5.1–15.5.2 ..... 15.6.5 ................. asabaliauskas on DSK5VPTVN1PROD with NOTICES 15.6.6 ................. 15.9A ................. 16.0 .................... 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. ML15125A386 ML15125A390 ML15126A149 ML15131A461 ML15125A307 ML15146A034 ML15131A245 ML15131A298 ML15131A305 ML15112A190 ML15084A403 ML15112A148 ML15126A086 ML15127A046 ML15131A329 ML15127A049 ML15127A349 ML15126A477 ML15131A417 ML15126A467 ML15118A943 ML15126A470 ML15131A261 ML15112A526 ML15124A607 ML15112A694 ML15119A057 ML15118A609 ML15125A367 ML15070A194 ML15070A204 ML15070A210 ML15084A407 ML15084A411 ML15125A294 ML15127A383 ML15127A373 ML15131A328 ML15125A297 ML15127A387 ML15127A391 ML15125A317 ML15127A395 ML15127A400 ML15125A292 ML15125A293 ML15118A927 ML15118A482 ML15118A600 ML15131A364 ML15118A474 ML15125A463 ML15131A334 ML15125A467 ML15131A311 ML15131A316 NUCLEAR REGULATORY COMMISSION [NRC–2015–0001] Sunshine Act Meeting Notice [FR Doc. 2015–16034 Filed 6–29–15; 8:45 am] June 29, July 6, 13, 20, 27, August 3, 2015. DATE: BILLING CODE 7590–01–P VerDate Sep<11>2014 17:34 Jun 29, 2015 Jkt 235001 PO 00000 Frm 00096 Fmt 4703 Sfmt 4703 E:\FR\FM\30JNN1.SGM 30JNN1

Agencies

NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 80, Number 125 (Tuesday, June 30, 2015)]
[Notices]
[Pages 37312-37314]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-16034]

[[Page 37312]]

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NUCLEAR REGULATORY COMMISSION

[NRC-2015-0160]

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

AGENCY: Nuclear Regulatory Commission.

ACTION: Design-specific review standard; request for comment.

-----------------------------------------------------------------------

SUMMARY: 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.

DATES: 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.

ADDRESSES: 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 https://www.regulations.gov and search for Docket ID NRC-2015-0160. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. 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.

FOR FURTHER INFORMATION CONTACT: Jenny Gallo, Office of New Reactors,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001;
telephone: 301-415-7367; email: NuScale-DSRS@nrc.gov.

SUPPLEMENTARY INFORMATION:

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 https://www.regulations.gov 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 https://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 (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 https://www.regulations.gov 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

[[Page 37313]]

NuScale 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.

----------------------------------------------------------------------------------------------------------------
Section Design-specific review standard title ADAMS Accession No.
----------------------------------------------------------------------------------------------------------------
Matrix........................... NuScale Power, LLC DSRS Scope and Safety ML15156B063
Review Matrix.
3.11............................. Environmental Qualification of Mechanical ML15131A247
and Electrical Equipment.
3.13............................. Threaded Fasteners--ASME Code Class 1, 2, ML15084A277
and 3.
3.3.1............................ Offsite Power System....................... ML15071A259
3.3.2............................ Tornado Loads.............................. ML15071A267
3.4.1............................ Internal Flood Protection for Onsite ML15139A112
Equipment Failures.
3.4.2............................ Analysis Procedures........................ ML15071A324
3.5.1.1.......................... Internally Generated Missiles (Outside ML15139A081
Containment).
3.5.1.2.......................... Internally Generated Missiles (Inside ML15139A096
Containment).
3.5.1.3.......................... Turbine Missiles........................... ML15070A248
3.5.1.4.......................... Missiles Generated by Tornadoes and Extreme ML15139A121
Winds.
3.5.2............................ Structures, Systems, and Components to be ML15139A102
Protected from Externally-Generated
Missiles.
3.5.3............................ Barrier Design Procedures.................. ML15071A273
3.7.1............................ Seismic Design Parameters.................. ML15084A279
3.7.2............................ Seismic System Analysis.................... ML15084A177
3.7.3............................ Seismic Subsystem Analysis................. ML15131A340
3.8.2............................ Steel Containment.......................... ML15131A373
3.8.4............................ Other Seismic Category I Structures........ ML15118A151
3.8.5............................ Foundations................................ ML15132A186
4.2.............................. Fuel System Design......................... ML15132A517
4.3.............................. Nuclear Design............................. ML15125A374
4.4.............................. Thermal and Hydraulic Design............... ML15131A427
4.5.2............................ Reactor Internal and Core Support Structure ML15070A325
Materials.
4.6.............................. Functional Design of Control Rod Drive ML15119A111
System.
5.2.2............................ Overpressure Protection.................... ML15118A931
5.2.4............................ Reactor Coolant Pressure Boundary Inservice ML15125A305
Inspection and Testing.
5.2.5............................ Reactor Coolant Pressure Boundary Leakage ML15132A194
Detection.
5.3.1............................ Reactor Vessel Materials................... ML15070A457
5.3.2............................ Pressure-Temperature Limits, ML15070A468
Upper[dash]Shelf Energy, and Pressurized
Thermal Shock.
5.3.3............................ Reactor Vessel Integrity................... ML15070A462
5.4.............................. Rx Coolant System Component and Subsystem ML15126A156
Design.
5.4.2.1.......................... Steam Generator Materials.................. ML15131A376
5.4.2.2.......................... Steam Generator Program.................... ML15070A562
5.4.7............................ Residual Heat Removal (RHR) System......... ML15131A360
5-4 BTP.......................... Design Requirements of the RHR System...... ML15132A524
6.1.1............................ Engineered Safety Features Materials....... ML15070A567
6.1.2............................ Protective Coating Systems (Paints)-- ML15071A372
Organic Materials.
6-1 BTP.......................... pH for Emergency Coolant Water for PWRs.... ML15125A369
6.2.1............................ Containment Functional Design.............. ML15118A922
6.2.1.1.A........................ PWR Dry Containments, Including Sub- ML15118A264
atmospheric Containments.
6.2.1.3.......................... Mass and Energy Release Analysis for ML15112A134
Postulated Loss-of-Coolant Accidents
(LOCAs).
6.2.1.4.......................... Mass and Energy Release Analysis for ML15118A293
Postulated Secondary System Pipe Ruptures.
6.2.2............................ Containment Heat Removal Systems........... ML15131A341
6.2.4............................ Containment Isolation System............... ML15119A087
6.2.5............................ Combustible Gas Control in Containment..... ML15119A090
6.2.6............................ Containment Leakage Testing................ ML15119A084
6.2.7............................ Fracture Prevention of Containment Pressure ML15112A517
Boundary.
6.3.............................. Emergency Core Cooling System.............. ML15125A322
6.6.............................. Inservice Inspection and Testing of Class 2 ML15127A136
and 3 Components.
7.0.............................. Instrumentation and Controls--Introduction ML15125A340
and Overview of Review Process.
7.0, A........................... Instrumentation and Controls--Hazard ML15132A583
Analysis.
7.0, B........................... Instrumentation and Controls--System ML15132A603
Architecture.
7.0, C........................... Instrumentation and Controls--Simplicity... ML15132A611
7.0, D........................... Instrumentation and Controls--References... ML15132A618
7.1.............................. I&C--Fundamental Design Principles......... ML15125A335
7.2.............................. Instrumentation and Controls--System ML15125A360
Characteristics.
8.1.............................. Electric Power--Introduction............... ML15146A269
8.2.............................. Offsite Power System....................... ML15125A425
8-2 BTP.......................... Use of Diesel-Generator Sets for Peaking... Ml15131A386
8.3.1............................ AC Power Systems (Onsite).................. ML15125A384

[[Page 37314]]

8.3.2............................ DC Power Systems (Onsite).................. ML15125A386
8-3 BTP.......................... Stability of Offsite Power Systems......... ML15125A390
8.4.............................. Station Blackout........................... ML15126A149
8-6 BTP.......................... Adequacy of Station Electric Distribution ML15131A461
System Voltages.
9.1.2............................ New and Spent Fuel Storage................. ML15125A307
9.1.3............................ Spent Fuel Pool Cooling and Cleanup System. ML15146A034
9.2.6............................ Condensate Storage Facilities.............. ML15131A245
9.3.2............................ Process and Post-Accident Sampling Systems. ML15131A298
9.3.4............................ Chemical and Volume Control System (PWR) ML15131A305
(Including Boron Recovery System).
9.3.6............................ Containment Evacuation and Flooding Systems ML15112A190
9.5.2............................ Communications Systems..................... ML15084A403
9.5.3............................ Lighting Systems........................... ML15112A148
10.2............................. Turbine Generator.......................... ML15126A086
10.2.3........................... Turbine Rotor Integrity.................... ML15127A046
10.3............................. Main Steam Supply System................... ML15131A329
10.4.1........................... Main Condensers............................ ML15127A049
10.4.2........................... Main Condenser Evacuation System........... ML15127A349
10.4.3........................... Turbine Gland Sealing System............... ML15126A477
10.4.4........................... Turbine Bypass System...................... ML15131A417
10.4.5........................... Circulating Water System................... ML15126A467
10.4.6........................... Condensate Cleanup System.................. ML15118A943
10.4.7........................... Condensate and Feedwater System............ ML15126A470
10.4.10.......................... Auxiliary Boiler System.................... ML15131A261
11.1............................. Source Terms............................... ML15112A526
11.2............................. Liquid Waste Management System............. ML15124A607
11.3............................. Gaseous Waste Management System............ ML15112A694
11.4............................. Solid Waste Management System.............. ML15119A057
11.5............................. Process and Effluent Radiological ML15118A609
Monitoring Instrumentation and Sampling
Systems.
11.6............................. Guidance on I&C Design Features for Process ML15125A367
and Effluent Radiological Monitoring and
Area Radiation and Airborne Radioactivity
Monitoring.
12.2............................. Radiation Sources.......................... ML15070A194
12.3-12.4........................ Radiation Protection Design Features....... ML15070A204
12.5............................. Operational Radiation Protection Program... ML15070A210
14.2............................. Initial Plant Test Program--Design ML15084A407
Certification and New License Applicants.
14.3.2........................... Structural and Systems Engineering-- ML15084A411
Inspections, Tests, Analyses, and
Acceptance Criteria.
14.3.4........................... Reactor Systems--Inspections, Tests, ML15125A294
Analyses, and Acceptance Criteria.
14.3.5........................... Instrumentation and Controls--Inspections, ML15127A383
Tests, Analyses, and Acceptance Criteria.
14.3.6........................... Electrical Systems--Inspections, Tests, ML15127A373
Analyses, and Acceptance Criteria.
14.3.7........................... Plant Systems--Inspections, Tests, ML15131A328
Analyses, and Acceptance Criteria.
15.0............................. Introduction--Transient and Accident ML15125A297
Analyses.
15.0.3........................... Design Basis Accidents Radiological ML15127A387
Consequence Analyses for Advanced Light
Water Reactors.
15.1.1-15.1.4.................... Decrease in FW Temperature, Increase in FW ML15127A391
Flow, Increase in Steam Flow, and
Inadvertent Opening of a Steam Generator
Relief or Safety Valve.
15.1.5........................... Steam System Piping Failures Inside and ML15125A317
Outside of Containment (PWR).
15.1.6........................... Loss of Containment Vacuum................. ML15127A395
15.2.1-15.2.5.................... Loss of External Load; Turbine Trip; Loss ML15127A400
of Condenser Vacuum; Closure of Main Steam
Isolation Valve (BWR); and Steam Pressure
Regulator Failure (Closed).
15.2.6........................... Loss of Non-Emergency AC Power to the ML15125A292
Station Auxiliaries.
15.2.7........................... Loss of Normal Feedwater Flow.............. ML15125A293
15.2.8........................... Feedwater System Pipe Breaks Inside and ML15118A927
Outside Containment (PWR).
15.4.1........................... Uncontrolled Control Rod Assembly ML15118A482
Withdrawal from a Subcritical or Low Power
Startup Condition.
15.4.2........................... Uncontrolled Control Rod Assembly ML15118A600
Withdrawal at Power.
15.4.3........................... Control Rod Misoperation (System ML15131A364
Malfunction or Operator Error).
15.4.6........................... Inadvertent Decrease in Boron Concentration ML15118A474
in the Reactor Coolant (PWR).
15.5.1-15.5.2.................... Chemical and Volume Control System ML15125A463
Malfunction that Increases Reactor Coolant
Inventory.
15.6.5........................... LOCAs Resulting From Spectrum of Postulated ML15131A334
Piping Breaks Within the Reactor Coolant
Pressure Boundary.
15.6.6........................... Inadvertent Opening of a PWR Pressurizer ML15125A467
Pressure Relief Valve.
15.9A............................ Thermal-hydraulic Stability................ ML15131A311
16.0............................. Technical Specifications................... ML15131A316
----------------------------------------------------------------------------------------------------------------

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.
[FR Doc. 2015-16034 Filed 6-29-15; 8:45 am]
BILLING CODE 7590-01-P

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