[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]
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