[Federal Register Volume 81, Number 210 (Monday, October 31, 2016)]
[Notices]
[Pages 75449-75452]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-26210]


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NUCLEAR REGULATORY COMMISSION

[NRC-2015-0160]


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

AGENCY: Nuclear Regulatory Commission.

ACTION: NuScale design-specific review standard; issuance.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC or Commission) has 
issued the NuScale Power, LLC, (NuScale), Design-Specific Review 
Standard (DSRS) Sections, and is issuing the final NuScale DSRS Scope 
and Safety Review Matrix, for NuScale Design Certification (DC), 
Combined License (COL), and Early Site Permit (ESP) reviews. The NRC 
staff is also issuing the DSRS public comment resolution matrices, 
which address the comments received on the draft DSRS. The NuScale DSRS 
provides guidance to the NRC staff for performing safety reviews for 
those specific areas where existing NUREG-0800, ``Standard Review Plan 
[SRP] for the Review of Safety Analysis Reports for Nuclear Power 
Plants: LWR Edition,'' sections do not address the unique features of 
the NuScale design.

DATES: The DSRS sections were effective upon issuance between June 24 
and August 4, 2016.

ADDRESSES: Please refer to Docket ID NRC-2015-0160 when contacting the 
NRC about the availability of information regarding this document. You 
may obtain publically-available information related to this document 
using any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0160. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection 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 [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document. The DSRS is available in ADAMS Package Accession No. 
ML15355A295 and the final NuScale DSRS Scope and Safety Review Matrix 
is also available in ADAMS under Accession No. ML16263A000. The 
resolution of

[[Page 75450]]

comments on the draft DSRS is documented in the DSRS Public Comment 
Resolution Matrices (ADAMS Package Accession No. ML16083A615). In 
addition, for the convenience of the reader, the ADAMS accession 
numbers are provided in a table in the ``Availability of Documents'' 
section of this document.
     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.

FOR FURTHER INFORMATION CONTACT: Rajender Auluck, telephone: 301-415-
1025; email: [email protected] or Gregory Cranston, telephone: 
301-415-0546; email: [email protected]; both are staff members 
of the Office of New Reactors, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001.

SUPPLEMENTARY INFORMATION: 

I. Background

    In the Staff Requirements Memorandum (SRM) COMGBJ-10-0004/COMGEA-
10-0001, ``Use of Risk Insights to Enhance 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.

II. DSRS for the NuScale Design

    The NuScale DSRS (available in ADAMS Package Accession No. 
ML15355A295) 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 Review 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 NuScale design; which SRP sections needed modification and were 
reissued as DSRS sections; and which new DSRS sections were added to 
address a unique design consideration in the NuScale design. The final 
NuScale DSRS Scope and Safety Review Matrix is available in ADAMS under 
Accession No. ML16263A000.
    The NRC staff developed the content of the NuScale DSRS as an 
alternative method for evaluating a NuScale-specific application and 
has determined that the application may address the DSRS in lieu of 
addressing the SRP, with specified exceptions. These exceptions include 
particular review areas in which the DSRS directs reviewers to consult 
the SRP and others in which the SRP is used for the review as 
identified in the final NuScale DSRS Scope and Safety Review Matrix. If 
NuScale chooses to address the DSRS, the application should identify 
and describe all differences between the design features (DC and COL 
applications only), analytical techniques, and procedural measures 
proposed in an application and the guidance of the applicable DSRS 
section (or SRP section, as specified in the NuScale DSRS Scope and 
Safety Review Matrix), and discuss how the proposed alternative 
provides an acceptable method of complying with the regulations that 
underlie the DSRS acceptance criteria. The staff has accepted the 
content of the DSRS as an alternative method for evaluating whether an 
application complies with NRC regulations for the NuScale Small Modular 
Reactor applications, provided that the application does not deviate 
significantly from the design and siting assumptions made by the NRC 
staff while preparing the DSRS. If the design or siting assumptions in 
a NuScale application deviate significantly from the design and siting 
assumptions the staff used in preparing the DSRS, the staff will use 
the more general guidance in the SRP, as specified in sections 
52.17(a)(1)(xii), 52.47(a)(9), or 52.79(a)(41) of title 10 of the Code 
of Federal Regulations, depending on the type of application. 
Alternatively, the staff may supplement the DSRS section by adding 
appropriate criteria to address new design or siting assumptions.
    The NRC staff issued a Federal Register notice on June 30, 2015 (80 
FR 37312), to request public comments on the draft NuScale DSRS Scope 
and Safety review Matrix (ADAMS Accession No. ML15156B063) and the 
individual NuScale-specific DSRS sections referenced in the table 
included in the FRN. A correction Federal Register notice was published 
on July 9, 2015 (80 FR 39454), to identify an additional draft DSRS 
section for which comments were requested. In response, the NRC 
received comments from: NuScale Power, LLC, by letter dated August 31, 
2015 (ADAMS Accession No. ML15258A081), the Nuclear Energy Institute 
(NEI) by letter dated August 31, 2015 (ADAMS Accession No. 
ML15257A012), Mark Thomson by electronic submission dated August 31, 
2015 (ADAMS Accession No. ML15292A309), an anonymous submitter by 
electronic submission dated August 31, 2015 (ADAMS Accession No. 
ML15292A310), an anonymous submitter by electronic submission dated 
August 31, 2015 (ADAMS Accession No. ML15292A311), Clinton Ferrara by 
electronic submission dated August 31, 2015 (ADAMS Accession No. 
ML15292A333), and Paula Ferrara by electronic submission dated August 
31, 2015, (ADAMS Accession No. ML15292A334). Several of these comments 
have been previously discussed during public meetings held in support 
of developing the draft DSRS sections. These comments and resolutions 
have been documented in the DSRS Public Comment Resolution Matrices and 
are publicly available (ADAMS Package Accession No. ML16083A615).
    In the June 30, 2015 Federal Register notice, the NRC requested 
public comments on 115 DSRS sections. The NRC staff determined whether 
to develop a DSRS section after considering whether significant 
differences in the functions, characteristics, or attributes of the 
NuScale design required major revision of the related SRP section 
guidance, or whether structures, systems, and components identified in 
the NuScale design are unique and not addressed by the current SRP. 
Following publication of the draft version of the DSRS sections, the 
NRC staff revisited these criteria and determined, based on the most 
recent NuScale design, that it is appropriate to use the related SRP 
section in lieu of a draft DSRS section

[[Page 75451]]

in a number of cases. In these cases the draft DSRS sections have not 
been issued as final, and the related SRP sections will be used for the 
NuScale review. In deciding to use the related SRP sections, the staff 
has not necessarily determined that the SRP sections are wholly 
applicable without modification. For example, as the NRC staff gains 
greater understanding of the NuScale design or if the design changes 
during the review, the staff would assess whether different or 
supplemental review criteria are needed. Stakeholders who believe that 
different or supplemental review criteria are needed may provide these 
views to the NRC staff for consideration during the application review 
period.
    The results of determinations to use the related SRP sections 
rather than draft DSRS sections, along with other identified issues 
with the draft NuScale DSRS Scope and Safety Review Matrix, are 
documented in a separate ``transitional'' NuScale DSRS Scope and Safety 
Review Matrix (ADAMS Accession No. ML16076A048). The ``transitional'' 
Matrix shows the differences between the draft and final NuScale DSRS 
Scope and Safety Review Matrices and describes the reasons for these 
differences. The resulting final list of DSRS titles with corresponding 
section numbers and ADAMS references are provided in the table below 
and in ADAMS Package Accession No. ML15355A295.
    In the future, should additional SRP sections be developed, the 
staff will determine at that time their applicability to the NuScale 
design. In addition, the NRC disseminates information regarding current 
safety issues and proposed solutions through various means, such as 
generic communications and the process for treating generic safety 
issues. When current issues are resolved, the staff will determine the 
need, extent, and nature of revision that should be made to the SRP 
and/or DSRS to reflect the new NRC guidance.

III. Availability of Documents

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                              Design-specific review    ADAMS accession
          Section                 standard title              No.
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Matrix....................  NuScale DSRS Scope and           ML16076A048
                             Safety Review Matrix
                             (Transitional).
Matrix....................  NuScale DSRS Scope and           ML16263A000
                             Safety Review Matrix
                             (Final).
3.5.1.3...................  Turbine Missiles.........        ML15355A364
3.7.1.....................  Seismic Design Parameters        ML15355A384
3.7.2.....................  Seismic System Analysis..        ML15355A389
3.7.3.....................  Seismic Subsystem                ML15355A402
                             Analysis.
3.8.2.....................  Steel Containment........        ML15355A411
3.8.4.....................  Other Seismic Category I         ML15355A444
                             Structures.
3.8.5.....................  Foundations..............        ML15355A451
3.11......................  Environmental                    ML15355A455
                             Qualification of
                             Mechanical and
                             Electrical Equipment.
4.4.......................  Thermal and Hydraulic            ML15355A468
                             Design.
5.2.4.....................  Reactor Coolant Pressure         ML15355A479
                             Boundary Inservice
                             Inspection and Testing.
5.2.5.....................  Reactor Coolant Pressure         ML15355A505
                             Boundary Leakage
                             Detection.
5.3.1.....................  Reactor Vessel Materials.        ML15355A513
5.3.2.....................  Pressure-Temperature             ML15355A526
                             Limits, Upper[dash]Shelf
                             Energy, and Pressurized
                             Thermal Shock.
5.3.3.....................  Reactor Vessel Integrity.        ML15355A530
5.4.2.1...................  Steam Generator Materials        ML15355A532
5.4.2.2...................  Steam Generator Program..        ML15355A535
5.4.7.....................  Decay Heat Removal (DHR)         ML15355A536
                             System.
BTP 5-4...................  Design Requirements of           ML15355A313
                             the Decay Heat Removal
                             System.
6.2.1.....................  Containment Functional           ML15356A259
                             Design.
6.2.1.1.A.................  Containments.............        ML15355A544
6.2.1.3...................  Mass and Energy Release          ML15357A327
                             Analysis for Postulated
                             Loss-of-Coolant
                             Accidents (LOCAs).
6.2.1.4...................  Mass and Energy Release          ML15356A241
                             Analysis for Postulated
                             Secondary System Pipe
                             Ruptures.
6.2.2.....................  Containment Heat Removal         ML15356A267
                             Systems.
6.2.4.....................  Containment Isolation            ML15356A332
                             System.
6.2.5.....................  Combustible Gas Control          ML15356A356
                             in Containment.
6.2.6.....................  Containment Leakage              ML15356A388
                             Testing.
6.3.......................  Emergency Core Cooling           ML15356A393
                             System.
6.6.......................  Inservice Inspection and         ML15356A396
                             Testing of Class 2 and 3
                             Components.
7.0.......................  Instrumentation and              ML15356A416
                             Controls--Introduction
                             and Overview of Review
                             Process.
7.1.......................  Instrumentation and              ML15363A293
                             Controls--Fundamental
                             Design Principles.
7.2.......................  Instrumentation and              ML15363A347
                             Controls--System
                             Characteristics.
7.0, App A................  Instrumentation and              ML15355A316
                             Controls--Hazard
                             Analysis.
7.0, App B................  Instrumentation and              ML15355A318
                             Controls--System
                             Architecture.
7.0, App C................  Instrumentation and              ML15355A319
                             Controls--Simplicity.
7.0, App D................  Instrumentation and              ML15355A320
                             Controls--References.
8.1.......................  Electric Power--                 ML15356A473
                             Introduction.
8.2.......................  Offsite Power System.....        ML15356A516
8.3.1.....................  AC Power Systems (Onsite)        ML15356A533
8.3.2.....................  DC Power Systems (Onsite)        ML15356A552
8.4.......................  Station Blackout.........        ML15356A570
9.1.2.....................  New and Spent Fuel               ML15356A584
                             Storage.
9.1.3.....................  Spent Fuel Pool Cooling          ML15356A595
                             and Cleanup System.
9.3.4.....................  Chemical and Volume              ML15356A622
                             Control System.
9.3.6.....................  Containment Evacuation           ML15356A637
                             and Flooding Systems.
9.5.2.....................  Communications Systems...        ML15363A400
10.2.3....................  Turbine Rotor Integrity..        ML15356A700
10.3......................  Main Steam Supply System.        ML15355A322
10.4.7....................  Condensate and Feedwater         ML15355A331
                             System.
11.1......................  Source Terms.............        ML15355A333

[[Page 75452]]

 
11.2......................  Liquid Waste Management          ML15355A334
                             System.
11.3......................  Gaseous Waste Management         ML15355A335
                             System.
11.4......................  Solid Waste Management           ML15355A336
                             System.
11.5......................  Process and Effluent             ML15355A337
                             Radiological Monitoring
                             Instrumentation and
                             Sampling Systems.
11.6......................  Guidance on                      ML15355A338
                             Instrumentation and
                             Control Design Features
                             for Process and Effluent
                             Radiological Monitoring,
                             and Area Radiation and
                             Airborne Radioactivity
                             Monitoring.
12.2......................  Radiation Sources........        ML15350A320
12.3-12.4.................  Radiation Protection             ML15350A339
                             Design Features.
12.5......................  Operational Radiation            ML15350A341
                             Protection Program.
14.2......................  Initial Plant Test               ML15355A339
                             Program--Design
                             Certification and New
                             License Applicants.
14.3.5....................  Instrumentation and              ML15355A340
                             Controls--Inspections,
                             Tests, Analyses, and
                             Acceptance Criteria.
15.0......................  Introduction--Transient          ML15355A302
                             and Accident Analyses.
15.0.3....................  Design Basis Accidents           ML15355A341
                             Radiological Consequence
                             Analyses for NuScale SMR
                             Design.
15.1.1--15.1.4............  Decrease in FW                   ML15355A303
                             Temperature, Increase in
                             FW Flow, Increase in
                             Steam Flow and
                             Inadvertent Opening of
                             the Turbine Bypass
                             System or Inadvertent
                             Operation of the Decay
                             Heat Removal System.
15.1.5....................  Steam System Piping              ML15355A304
                             Failures Inside and
                             Outside of Containment.
15.1.6....................  Loss of Containment              ML15355A305
                             Vacuum.
15.2.1-15.2.5.............  Loss of External Load;           ML15355A306
                             Turbine Trip; Loss of
                             Condenser Vacuum;
                             Closure of Main Steam
                             Isolation Valve (BWR);
                             and Steam Pressure
                             Regulator Failure
                             (Closed).
15.2.6....................  Loss of Non-Emergency AC         ML15363A348
                             Power to the Station
                             Auxiliaries.
15.2.7....................  Loss of Normal Feedwater         ML15355A307
                             Flow.
15.2.8....................  Feedwater System Pipe            ML15355A308
                             Breaks Inside and
                             Outside Containment.
15.5.1-15.5.2.............  Chemical and Volume              ML15363A397
                             Control System
                             Malfunction that
                             Increases Reactor
                             Coolant Inventory.
15.6.5....................  LOCAs Resulting From             ML15355A309
                             Spectrum of Postulated
                             Piping Breaks Within the
                             Reactor Coolant Pressure
                             Boundary.
15.6.6....................  Inadvertent Opening of           ML15355A310
                             the Emergency Core
                             Cooling System.
15.9A.....................  Thermal-hydraulic                ML15355A311
                             Stability.
16.0......................  Technical Specifications.        ML15355A312
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    Dated at Rockville, Maryland, this 21st day of October 2016.

    For the Nuclear Regulatory Commission.
Frank Akstulewicz,
Director, Division of New Reactor Licensing, Office of New Reactors.
[FR Doc. 2016-26210 Filed 10-28-16; 8:45 am]
 BILLING CODE 7590-01-P