[Federal Register Volume 88, Number 12 (Thursday, January 19, 2023)]
[Rules and Regulations]
[Pages 3287-3311]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2023-00729]
========================================================================
Rules and Regulations
Federal Register
________________________________________________________________________
This section of the FEDERAL REGISTER contains regulatory documents
having general applicability and legal effect, most of which are keyed
to and codified in the Code of Federal Regulations, which is published
under 50 titles pursuant to 44 U.S.C. 1510.
The Code of Federal Regulations is sold by the Superintendent of Documents.
========================================================================
Federal Register / Vol. 88, No. 12 / Thursday, January 19, 2023 /
Rules and Regulations
[[Page 3287]]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 52
[NRC-2017-0029]
RIN 3150-AJ98
NuScale Small Modular Reactor Design Certification
AGENCY: U.S. Nuclear Regulatory Commission.
ACTION: Final rule.
-----------------------------------------------------------------------
SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its
regulations to certify the NuScale standard design for a small modular
reactor. Applicants or licensees intending to construct and operate a
NuScale standard design may do so by referencing this design
certification rule. The applicant for certification of the NuScale
standard design is NuScale Power, LLC.
DATES: This final rule is effective on February 21, 2023. The
incorporation by reference of certain publications listed in the rule
is approved by the Director of the Federal Register as of February 21,
2023.
ADDRESSES: Please refer to Docket ID NRC-2017-0029 when contacting the
NRC about the availability of information for this action. You may
obtain publicly available information related to this action by any of
the following methods:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0029. Address
questions about NRC dockets to Dawn Forder; telephone: 301-415-3407;
email: [email protected]. For technical questions, contact the
individuals 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 https://www.nrc.gov/reading-rm/adams.html. To begin the search, 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]. For the convenience of the reader,
instructions about obtaining materials referenced in this document are
provided in the ``Availability of Documents'' section.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room P1 B35, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852. To make an appointment to
visit the PDR, please send an email to [email protected] or call 1-
800-397-4209 or 301-415-4737, between 8:00 a.m. and 4:00 p.m. (ET),
Monday through Friday, except Federal holidays.
Technical Library: The Technical Library, which is located
at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland
20852, is open by appointment only. Interested parties may make
appointments to examine documents by contacting the NRC Technical
Library by email at [email protected] between 8:00 a.m. and 4:00
p.m. (ET), Monday through Friday, except Federal holidays.
FOR FURTHER INFORMATION CONTACT: Yanely Malave, Office of Nuclear
Material Safety and Safeguards, telephone: 301-415-1519, email:
[email protected], and Carolyn Lauron, Office of Nuclear Reactor
Regulation, telephone: 301-415-2736, email: [email protected].
Both are staff of the U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
SUPPLEMENTARY INFORMATION:
Table of Contents
I. Background
II. Opportunities for Public Participation
III. Regulatory and Policy Issues
IV. Technical Issues Associated With the NuScale Design
V. Discussion
A. Introduction (Section I)
B. Definitions (Section II)
C. Scope and Contents (Section III)
D. Additional Requirements and Restrictions (Section IV)
E. Applicable Regulations (Section V)
F. Issue Resolution (Section VI)
G. Duration of This Appendix (Section VII)
H. Processes for Changes and Departures (Section VIII)
I. [Reserved] (Section IX)
J. Records and Reporting (Section X)
VI. Public Comment Analysis
VII. Section-by-Section Analysis
VIII. Regulatory Flexibility Certification
IX. Regulatory Analysis
X. Backfitting and Issue Finality
XI. Plain Writing
XII. Environmental Assessment and Finding of No Significant Impact
XIII. Paperwork Reduction Act
XIV. Congressional Review Act
XV. Agreement State Compatibility
XVI. Voluntary Consensus Standards
XVII. Availability of Documents
XVIII. Incorporation by Reference--Reasonable Availability to
Interested Parties
I. Background
Part 52 of title 10 of the Code of Federal Regulations (10 CFR),
``Licenses, Certifications, and Approvals for Nuclear Power Plants,''
subpart B, ``Standard Design Certifications,'' presents the process for
obtaining standard design certifications. By letter dated December 31,
2016, NuScale Power, LLC, (NuScale Power) filed its application for
certification of the NuScale standard design (hereafter referred to as
NuScale). The NRC published a notification of receipt of the design
certification application (DCA) in the Federal Register on February 22,
2017 (82 FR 11372). On March 30, 2017, the NRC published a notification
of acceptance for docketing of the application in the Federal Register
(82 FR 15717) and assigned docket number 52-048. The preapplication
information submitted before the NRC formally accepted the application
can be found in ADAMS under Docket No. PROJ0769.
NuScale is the first small modular reactor design reviewed by the
NRC. NuScale is based on a small light water reactor developed at
Oregon State University in the early 2000s. It consists of one or more
NuScale power modules (hereafter referred to as power module(s)). A
power module is a natural circulation light water reactor composed of a
reactor core, a pressurizer, and two helical coil steam generators
located in a common reactor pressure vessel that is housed in a compact
cylindrical steel containment. The NuScale reactor building is designed
to hold up to 12 power modules. Each power module has a rated thermal
output of 160 megawatt thermal (MWt) and electrical output of 50
megawatt electric (MWe), yielding a
[[Page 3288]]
total capacity of 600 MWe for 12 power modules. All the NuScale power
modules are partially submerged in a common safety-related pool, which
is also the ultimate heat sink for up to 12 power modules. The pool
portion of the reactor building is located below grade. The design
utilizes several first-of-a-kind approaches for accomplishing key
safety functions, resulting in no need for Class 1E safety-related
power (no emergency diesel generators), no need for pumps to inject
water into the core for post-accident coolant injection, and reduced
need for control room staffing while providing safe operation of the
plant during normal and post-accident operation.
II. Opportunities for Public Participation
The proposed rule and environmental assessment were published in
the Federal Register on July 1, 2021, for a 60-day public comment
period (86 FR 34999). The public comment period was scheduled to close
on August 30, 2021. The NRC subsequently extended the comment period by
45 days (86 FR 47251; August 24, 2021), providing a total comment
period of 105 days. The public comment period closed on October 14,
2021. The public comments informed the development of this final rule.
III. Regulatory and Policy Issues
A. Exemptions for Future Applicants Referencing NuScale
1. Control Room Staffing Requirements
The requirements in Sec. Sec. 50.54(k) and 50.54(m) identify the
minimum number of licensed operators that must be on site, in the
control room, and at the controls. The requirements are conditions in
every nuclear power reactor operating license issued under 10 CFR part
50, ``Domestic Licensing of Production and Utilization Facilities.''
The requirements also are conditions in every combined license (COL)
issued under 10 CFR part 52; however, they are applicable only after
the Commission makes the finding under Sec. 52.103(g) that the
acceptance criteria in the COL are met.
In a letter to the NRC, dated September 15, 2015, NuScale Power
proposed that 6 licensed operators would operate up to 12 power modules
from a single control room. The staffing proposal would meet the
requirements of Sec. 50.54(k) but would not meet the requirements in
Sec. 50.54(m)(2)(i) because the minimum requirements for the onsite
staffing table in Sec. 50.54(m)(2)(i) do not address operation of more
than two units from a single control room. The proposal also would not
meet Sec. 50.54(m)(2)(iii), which requires a licensed operator at the
controls for each fueled unit. Absent alternative staffing
requirements, future applicants referencing the NuScale design would
need to request an exemption.
In DCA, Part 7, Section 6, NuScale requested that the NRC approve
design-specific control room staffing requirements in lieu of the
requirements in Sec. 50.54(m). In the DCA Part 7, Section 6.2,
``Justification for Rulemaking,'' NuScale Power provided a technical
basis for its proposed alternative control room staffing requirements.
NuScale Power's proposed approach is consistent with SECY-11-0098,
``Operator Staffing for Small or Multi-Module Nuclear Power Plant
Facilities,'' dated July 22, 2011. For the reasons described in Chapter
18, Section 18.5.4.2, ``Evaluation of the Applicant's Technical
Basis,'' of the final safety evaluation report, the NRC found that
NuScale Power's proposed staffing level, as described in the DCA Part
7, Section 6, is acceptable. Because Section V, ``Applicable
Regulations,'' of this final rule includes the alternative staffing
requirement provisions, staffing table, and appropriate table notes, a
future applicant or licensee that references appendix G to 10 CFR part
52 will not need to request an exemption from Sec. 50.54(m).
2. Preoperational and Periodic Testing of Primary Reactor Containment
General Design Criterion (GDC) 52, ``Capability for Containment
Leakage Rate Testing,'' requires that the containment be designed so
that periodic, integrated leakage rate testing can be conducted at
containment design pressure; the underlying purpose of which is to
provide design capability for testing that assures that containment
leakage integrity is maintained and containment vessel leakage does not
exceed allowable leakage rate values (see appendix J to 10 CFR part
50). Under 10 CFR 50.54(o), operating licenses and combined licenses
for certain water-cooled power reactors must include a condition that
the primary containment shall be subject to appendix J to 10 CFR part
50, ``Primary Reactor Containment Leakage Testing for Water-Cooled
Power Reactors.'' Appendix J to 10 CFR part 50 requires that primary
reactor containments meet the containment leakage test requirements to
provide for preoperational and periodic verification by tests of the
leak-tight integrity of the primary reactor containment (Type A) and
systems and components that penetrate containment (Type B and Type C).
NuScale Power requested an exemption from GDC 52 in order to not
design NuScale to include the capability for Type A testing and
requested that the design certification rule exempt licensees
referencing the NuScale design certification rule from the requirement
for Type A testing in appendix J to 10 CFR part 50. NuScale Power's
request was based on the NuScale small modular reactor design meeting
the underlying purpose of the regulation through means not anticipated
when the NRC issued GDC 52 and appendix J to 10 CFR part 50. NuScale
Power stated that the NuScale containment has two primary features
distinguishing it from containments at existing light water reactors
that provide assurance that no unknown leakage pathways will be
present. First, the NuScale containment is designed and would be
constructed as a pressure vessel, and therefore leakage due to vessel
design or fabrication flaws would be identified during a required
preservice structural integrity test. In contrast to a Type A test,
this test is a hydrostatic leakage test at design pressure, with no
visible leakage as its acceptance criterion. Second, the containment is
100-percent inspectable, both inside and outside, whereby aging-related
flaws leading to potential leakage could be observed. Containment
leakage integrity assurance for NuScale is described in detail in
technical report TR-1116-51962-NP, ``NuScale Containment Leakage
Integrity Assurance,'' Rev. 1 (May 2019), which this final rule
incorporates by reference. NuScale Power stated that the required
preservice tests and inservice inspections described in TR-1116-51962-
NP, including Type B and Type C testing without Type A testing, ensure
that containment leakage rates remain acceptable.
In Chapter 6, Section 6.2.6.4, ``Technical Evaluation for Exemption
Request No. 7,'' of the final safety evaluation report, the NRC staff
concluded that granting this exemption from Type A testing, and
associated design features required by GDC 52 to provide for Type A
testing, is acceptable because the NuScale design relies on the
preservice pressure test, successful Type B and C testing at each
refueling as required in appendix J to 10 CFR part 50, periodic
inservice inspections, and direct observation of the entire vessel to
identify potential degradation or unknown leakage pathways for the
remainder of the service life for the containment.
[[Page 3289]]
The NRC received a comment that the exemption from the requirement
for Type A testing in appendix J to 10 CFR part 50 should have been
listed in the proposed rule. The NRC agrees that the exemption should
have been included in the proposed rule. The NRC's conclusion that Type
A testing is not necessary for NuScale was noticed for comment as the
basis for the exemption from GDC 52. The exemption from Type A testing
itself was discussed in detail in the same section of final safety
evaluation report that evaluated the exemption from GDC 52. Although
the exemption from Type A testing was not included in the proposed
rule, the change to this final rule only specifies that future
licensees that reference this final rule will not be required to
perform Type A testing for which NuScale is not designed or required to
be capable of. Therefore, the NRC concludes that the exemption from the
Type A test in appendix J to 10 CFR part 50 is a logical outgrowth of
the proposed rule. In addition, because the issue of whether Type A
testing is necessary for NuScale was noticed in the proposed rule and
the NRC received no comments on the matter, the NRC finds that notice
and comment on this exemption from Type A testing is unnecessary within
the meaning of 5 U.S.C. 553(b).
Thus, Section V, ``Applicable Regulations,'' in this final rule
includes an exemption for licensees referencing appendix G to 10 CFR
part 52 from the requirement of appendix J to 10 CFR part 50 to conduct
Type A testing.
B. Incorporation by Reference
Section III.A, ``Incorporation by reference approval,'' of appendix
G to 10 CFR part 52 lists documents that were approved by the Director
of the Office of the Federal Register for incorporation by reference
into this appendix. Section III.B.2 identifies information that is not
within the scope of the design certification and, therefore, is not
incorporated by reference into this appendix. This information includes
conceptual design information, as defined in Sec. 52.47(a)(24), and
the discussion of ``first principles'' described in the Design Control
Document (DCD) Part 2, Tier 2, Section 14.3.2, ``Tier 1 Design
Description and Inspections, Tests, Analyses, and Acceptance Criteria
First Principles.''
The final rule has been updated to align with the Office of the
Federal Register's latest guidance for incorporation by reference,
issued on March 1, 2022, as supplemented by Release 1-2022 to the
Incorporation by Reference Handbook.
C. Issues Not Resolved by the Design Certification
The NRC identified three issues as not resolved within the meaning
of Sec. 52.63(a)(5). There was insufficient information available for
the NRC to resolve issues regarding (1) the shielding wall design in
certain areas of the plant, (2) the potential for containment leakage
from the combustible gas monitoring system, and (3) the ability of the
steam generator tubes to maintain structural and leakage integrity
during density wave oscillations in the secondary fluid system,
including the method of analysis to predict the thermal-hydraulic
conditions of the steam generator secondary fluid system and resulting
loads, stresses, and deformations from density wave oscillations from
reverse flow.
1. Shielding Wall Design
As discussed in Section 12.3.4.1.2 of the final safety evaluation
report, the NRC found that there were insufficient design details
available regarding shielding wall design with the presence of large
penetrations, such as the main steam lines; main feedwater lines; and
power module bay heating, ventilation, and air conditioning lines in
the radiation shield wall between the power module bay and the reactor
building steam gallery area. Without this shielding design information,
the NRC is unable to confirm that the radiological doses to workers
will be maintained within the radiation zone limits specified in the
application.
This issue is narrowly focused on the shielding walls between the
reactor module bays and the reactor building steam gallery areas. The
radiation zones and dose calculations, including dose calculations for
the dose to workers, members of the public, and environmental
qualification, in areas outside of the reactor module bay are
calculated assuming a solid wall and currently do not account for
penetrations in the shield wall. An applicant is required to
demonstrate penetration shielding adequate to address the following
issues in the NuScale DCD: the plant radiation zones, environmental
qualification dose calculations, and dose estimates for workers and the
public. An applicant can provide this information for the NRC to review
because this issue involves a localized area of the plant without
affecting other aspects of the NRC's review of the NuScale design.
Therefore, the NRC has determined that this information can be provided
by an applicant that references this appendix without a demonstrable
impact on safety or standardization. Appendix G to 10 CFR part 52,
Section VI, ``Issue Resolution,'' clarifies that this issue is not
resolved within the meaning of Sec. 52.63(a)(5), and Section IV,
``Additional Requirements and Restrictions,'' states that the COL
applicant is responsible for providing the design information to
address this issue.
2. Containment Leakage From the Combustible Gas Monitoring System
As documented in Section 12.3.4.1.3 of the final safety evaluation
report, there was insufficient information available regarding the
NuScale combustible gas monitoring system and the potential for leakage
from this system outside containment. Without additional information
regarding the potential for leakage from this system, the NRC was
unable to determine whether this leakage could impact analyses
performed to assess main control room dose consequences, offsite dose
consequences to members of the public, and whether this system can be
safely re-isolated after monitoring is initiated due to potentially
high dose levels at or near the isolation valve location. The isolation
valve can only be operated locally, and dose levels at the valve
location have not been determined.
This issue is narrowly focused on the radiation dose implications
as a result of using the post-accident combustible gas monitoring loop.
An applicant is required under Sec. Sec. 50.34(f)(2) and 52.47(a)(2)
to demonstrate either that offsite and main control room dose
calculations are not exceeded or that the system can be safely re-
isolated, if needed. This issue does not affect normal plant operation
or non-core damage accidents. The issue may be resolved by performing
radiation dose calculations and demonstrating that doses would remain
within applicable dose limits in 10 CFR part 20, ``Standards for
Protection Against Radiation.'' More information may be available at
the application stage that would allow for more detailed calculations.
Any design changes to address this issue would only affect the
combustible gas monitoring loop to ensure it can be re-isolated or to
ensure that dose limits are not exceeded. Such design changes likely
would not have an impact on other systems or equipment, and the NRC
would review such changes and any resulting effects on other
structures, systems, and components during the application review to
determine whether there is reasonable assurance of adequate
[[Page 3290]]
protection of public health and safety. Therefore, the NRC has
determined that this information can be provided by an applicant that
references this appendix without a demonstrable impact on safety or
standardization. Appendix G to 10 CFR part 52, Section VI, ``Issue
Resolution,'' clarifies that this issue is not resolved within the
meaning of Sec. 52.63(a)(5), and Section IV, ``Additional Requirements
and Restrictions,'' states that the COL applicant is responsible for
providing the design information to address this issue.
3. Steam Generator Stability During Density Wave Oscillations and
Associated Method of Analysis
Section 5.4.1.2, ``System Design,'' in Revision 2 of the DCA Part
2, Tier 2 (ADAMS Accession No. ML18310A345), stated that a flow
restriction device at the inlet to each steam generator tube ``ensures
secondary-side flow stability and precludes density wave
oscillations.'' However, the applicant modified this section in
Revision 3 of the DCA Part 2, Tier 2 (ADAMS Accession No. ML19241A431),
to state that the steam generator inlet flow restrictors provide the
necessary secondary-side pressure drop ``to reduce flow oscillations to
acceptable limits.'' Revision 4.1 of the DCA (ADAMS Accession No.
ML20205L562) revised Section 5.4.1.2 to state that the steam generator
inlet flow restrictors are designed ``to reduce the potential for
density wave oscillations.'' Revision 5 of this section of the DCA
(ADAMS Accession No. ML20225A071) provides only editorial changes to
Revision 4.1 and does not change the technical content or conclusions.
Sections 3.9.2, 3.9.5, and 5.4.1 of the final safety evaluation
report relied on the applicant's statements in Revision 2 and Revision
3 of the DCA that flow oscillations in the secondary fluid system of
the steam generators would either be precluded or minimal. After
issuance of the advanced safety evaluation report, the NRC noted
inconsistencies and gaps in the information provided in Sections 3.9.1,
3.9.2, and 5.4.1 of Revision 4.1 of the DCA Part 2, Tier 2, regarding
the potential for significant density wave oscillations in the steam
generator tubes, including both forward and reverse secondary flow. The
testing performed by the applicant on various conceptual designs of the
steam generator inlet flow restrictors only involved flow in the
forward direction without oscillation or reverse flow.
As a result, NuScale Power has not demonstrated that the flow
oscillations that are predicted to occur on the secondary side of the
steam generators will not cause failure of the inlet flow restrictors.
Structural and leakage integrity of the inlet flow restrictors in the
steam generators is necessary to avoid damage to multiple steam
generator tubes, caused directly by broken parts or indirectly by
unexpected density wave oscillation loads. Damage to multiple steam
generator tubes could disrupt natural circulation in the reactor
coolant pathway and interfere with the decay heat removal system and
the emergency core cooling system, which is relied upon to cool the
reactor core in a NuScale power module. The failure of multiple steam
generator tubes resulting from failure of an inlet flow restrictor has
not been included within the scope of the NuScale accident analyses in
DCA Part 2, Tier 2, Chapter 15. Therefore, the NRC concludes that
NuScale Power has not demonstrated compliance with 10 CFR
52.47(a)(2)(iv) and appendix A to 10 CFR part 50, GDC 4 and GDC 31,
relative to potential impacts on steam generator tube integrity from
inlet flow restrictor failure.
As described previously, NuScale Power made a change to the
description of inlet flow restrictor performance beginning with DCA
Part 2, Tier 2, Revision 3, that indicates that the design no longer
precludes density wave oscillations in the secondary side of the steam
generators. As a result, the design needs a method of analysis to
predict the thermal-hydraulic conditions of the steam generator
secondary fluid system and resulting loads, stresses, and deformations
from density wave oscillations including reverse flow. However, as
described in the next paragraph, NuScale power did not provide
verification and validation for its proposed method of analysis to
demonstrate it is appropriate for this purpose.
The DCA Part 2, Tier 2, Section 3.9.1.2, ``Computer Programs Used
in Analyses,'' lists the computer programs used by NuScale Power in the
dynamic and static analyses of mechanical loads, stresses, and
deformations, and in the hydraulic transient load analyses of seismic
Category I components and supports for the NuScale nuclear power plant.
Section 3.9.1.2 states that NRELAP5 is NuScale's proprietary system
thermal-hydraulics code for use in safety-related design and analysis
calculations and is pre-verified and configuration-managed. The
advanced safety evaluation report, Section 3.9.1.4.9, ``Computer
Programs Used in Analyses,'' states that the NRELAP5 computer program
had received verification and validation. Following preparation of the
advanced safety evaluation report, the NRC noted a discrepancy between
two statements in the DCA about validation for NRELAP5: DCA Part 2,
Tier 2, Section 5.4.1.3, in Revision 4 stated that NRELAP5 was
validated for determining density wave oscillation thermal-hydraulic
conditions, referring to Section 15.0.2 for more information, but
neither Section 15.0.2 nor technical report TR-1016-51669-NP describe
validation for determining density wave oscillation thermal-hydraulic
conditions.
On June 19, 2020, NuScale submitted Revision 4.1 of the DCA Part 2,
Tier 2 (ADAMS Accession No. ML20205L562; subsequently included in
Revision 5 of the DCA submitted on July 29, 2020 (ADAMS Accession No.
ML20225A071)), to correct the discrepancies and acknowledge the need
for a COL applicant to address secondary-side instabilities in the
steam generator design. Specifically, the update to Section 3.9.1.2 in
Revision 4.1 of DCA Part 2, Tier 2, references DCA Part 2, Tier 2,
Section 15.0.2, ``Review of Transient and Accident Analysis Methods,''
for the discussion of the development, use, verification, validation,
and code limitations of the NRELAP5 computer program for application to
transient and accident analyses. The correction to Section 3.9.1.2 also
references technical report TR-1016-51669-NP, ``NuScale Power Module
Short-Term Transient Analysis,'' incorporated by reference in DCA Part
2, Tier 2, Table 1.6-2, for application of the NRELAP5 computer program
to short-term transient dynamic mechanical loads, such as pipe breaks
and valve actuations. In addition, the correction to Section 3.9.1.2
includes a new COL item specifying that a COL applicant that references
the NuScale DCD will develop an evaluation methodology for the analysis
of secondary-side instabilities in the steam generator design. The COL
item states that this methodology would address the identification of
potential density wave oscillations in the steam generator tubes and
qualification of the applicable portions of the reactor coolant system
integral reactor pressure vessel and steam generator given the
occurrence of density wave oscillations, including the effects of
reverse fluid flows within the tubes. These corrections to the DCA
clarify that the evaluation methodology for the analysis of secondary-
side instabilities in the steam generator design was not verified and
validated as
[[Page 3291]]
part of the NuScale DCA but will need to be established by the COL
applicant.
This steam generator design issue is narrowly focused on the
effects of density wave oscillations in the secondary fluid system on
steam generator tubes to maintain structural and leakage integrity,
including the method of analysis to predict the thermal-hydraulic
conditions of the steam generator secondary fluid system and resulting
loads, stresses, and deformations from density wave oscillations
including reverse flow. No other reactor safety aspect of the steam
generators is impacted by this design issue. As a result, the NRC finds
that this is an isolated issue that does not affect other aspects of
the NRC's review of the design of the NuScale nuclear power plant.
Therefore, the NRC has determined that this information can be provided
by an applicant that references this appendix, consistent with the
other design information regarding steam generator integrity described
in DCA Part 2, Tier 2, Sections 3.9.1, 3.9.2, and 5.4.1, without a
demonstrable impact on safety or standardization. Therefore, appendix G
to 10 CFR part 52, Section VI, ``Issue Resolution,'' clarifies that
this issue is not resolved within the meaning of Sec. 52.63(a)(5), and
Section IV, ``Additional Requirements and Restrictions,'' states that
the COL applicant is responsible for providing the design information
to address this issue.
D. The Term ``Multi-Unit'' as Applied to NuScale
In a letter response to NuScale Power dated October 25, 2016, the
NRC staff explained how the staff's review of NuScale would apply the
definitions for ``nuclear power unit'' from Appendix A to 10 CFR part
50, ``General Design Criteria for Nuclear Power Plants,'' and ``modular
design'' from Sec. 52.1, ``Definitions.'' As defined in Appendix A to
10 CFR part 50, a nuclear power unit is the combination of a nuclear
reactor and the equipment for power generation. As defined in Sec.
52.1, modular design means that the nuclear power station consists of
two or more essentially identical nuclear reactors (modules) and that
each module is capable of operation independent of the other modules,
even if they have some shared systems.
The NuScale modular design combines one or more nuclear reactors
(up to 12) with the necessary equipment for power generation, such that
each separate nuclear reactor can be operated independent of the stage
of completion or operating condition of any other nuclear reactor on
the same site. Therefore, each reactor (i.e., power module) is a
separate nuclear power unit. However, NuScale's modular design means
that some multi-unit considerations are integral to the design. The
NuScale DCD addresses multi-unit considerations other than construction
for up to 12 power modules in a single reactor building, but the
NuScale DCD does not address multi-unit issues that may arise if a
NuScale facility is constructed and operated on the same site as
another nuclear facility.
For previously certified or licensed power reactor designs (one
nuclear power unit per reactor building), multi-unit site
considerations arose when multiple nuclear power units (in separate
reactor buildings) on the same site could affect the construction or
operation of another unit in a manner not previously reviewed by the
NRC. However, because the NuScale design has been reviewed and is
certified for multiple units in a single reactor building, issues
related to multiple NuScale units in the same reactor building
constructed at the same time have been resolved. Future applicants
referencing the NuScale design certification will need to address
multi-unit construction issues and, if applicable, multi-unit issues
for a proposed NuScale facility to be constructed and operated on the
same site as another nuclear facility, including adding additional
NuScale modules to a previously licensed NuScale reactor building.
The NRC has added a definition of the term ``nuclear power unit''
to this final rule.
IV. Technical Issues Associated With the NuScale Design
The NRC identified significant technical issues associated with the
following design areas that were resolved during the review:
Comprehensive vibration assessment program;
Containment safety analysis;
Emergency core cooling system inadvertent actuation block
valve;
Conformance with GDC 27, ``Combined Reactivity Control
Systems Capability,'' of appendix A, ``General Design Criteria for
Nuclear Power Plants,'' to 10 CFR part 50;
Absence of safety-related Class 1E alternating current
(AC) or direct current (DC) electrical power;
Accident source term methodology;
Boron redistribution during passive cooling modes.
In addition, the NRC granted 17 exemptions from 10 CFR part 50 to
address various aspects of NuScale Power's design.
A. Comprehensive Vibration Assessment Program
The NuScale comprehensive vibration assessment program limits
potentially adverse effects from flow, acoustic, and mechanically
induced vibrations and resonances on NuScale power module components,
including the helical coil steam generators. The NuScale steam
generators are different from those of operating pressurized-water
reactors in that the primary reactor coolant is on the outside of the
steam generator tubes and the steam is on the inside. Because of this
design, there is the possibility of density wave oscillation
instabilities in the secondary coolant, which could challenge the
integrity of the tubes. The NRC's review and findings, including
independent analyses and observation of vibration testing, are
documented in detail in Chapter 3, ``Design of Structures, Systems,
Components and Equipment,'' Section 3.9.2, ``Dynamic Testing and
Analysis of Systems, Structures, and Components,'' of the final safety
evaluation report. The review focused on assuring that the design of
the helical coil steam generator tubes would not result in issues with
flow-induced vibration.
As part of the comprehensive vibration assessment, the NRC also
reviewed and found acceptable the steam generator tube margin against
fluid-elastic instability, steam generator tube margin against vortex
shedding, control rod drive shaft margin against vortex shedding, in-
core instrument guide tube against vortex shedding, decay heat removal
system piping against acoustic resonance, and control rod assembly
guide tube against turbulence buffeting. The steam generator tube
margins against fluid-elastic instability and vortex shedding will be
validated in the TF-3 testing facility as described in DCA Part 2, Tier
1, Section 2.1.1, ``Design Description.'' In addition, the initial
startup testing will confirm that flow-induced vibration will not cause
adverse effects on the plant system components including the steam
generator tubes. With the exception of the steam generator tube and
inlet flow restrictor issue discussed in Section III.C.3, the NRC found
the comprehensive vibration assessment program adequate to ensure the
structural integrity of the NuScale power module components.
B. Containment Safety Analysis
NuScale incorporates novel and unique features that result in
transient thermal-hydraulic responses that are different from those of
currently licensed reactors.
[[Page 3292]]
There are several peak containment pressure analysis technical
issues unique to NuScale, including the associated thermal-hydraulic
analyses. In support of containment safety analysis, NuScale Power
submitted technical report TR-0516-49084-NP, Revision 3, ``Containment
Response Analysis Methodology,'' May 2020, which describes the
conservative containment pressure and temperature safety analyses for
several design-basis events related to the containment design margins.
NuScale Power also submitted topical report TR-0516-49422-NP, ``Loss-
of-Coolant Accident Evaluation Model,'' Revision 1, dated November
2019. This topical report describes the evaluation model used to
analyze the power module response during a design-basis loss-of-coolant
accident. The NRC reviewed this topical report as part of the
containment safety analysis.
The NRC also observed thermal-hydraulic performance testing at
NuScale Power's integrated system test facility, which validates the
analytical model. Based on initial testing results and thermal-
hydraulic analyses, NuScale Power made design changes to increase the
initial reactor building pool level and the in-containment vessel
design pressure to account for some uncertainties.
The NRC reviewed the details of the computer thermal-hydraulic
evaluation model described in the DCA Part 2, Tier 2, Section 6.2.1.1,
to determine whether any uncertainties were properly accounted for and
found the containment design margins to be acceptable. The associated
safety evaluation report approving topical report TR-0516-49422 was
issued on February 18, 2020. The NRC's review and specific findings,
including independent analyses and observation of NuScale testing, are
documented in Chapter 6, ``Engineered Safety Features,'' Section
6.2.1.1, ``Containment Structure,'' of the safety evaluation report.
C. Emergency Core Cooling System Inadvertent Actuation Block Valve
The NuScale emergency core cooling system relies on natural
circulation cooling of the reactor core by releasing the heated reactor
coolant steam from the top of the reactor pressure vessel through three
reactor vent valves into the containment vessel and returning the
cooled condensed reactor coolant water to the reactor pressure vessel
through two reactor recirculation valves. Each reactor vent valve and
reactor recirculation valve consists of a first-of-a-kind arrangement
of a main valve, an inadvertent actuation block (IAB) valve, a solenoid
trip valve, and a solenoid reset valve. The IAB valve for each reactor
vent valve and reactor recirculation valve is designed to close rapidly
to prevent its corresponding emergency core cooling system main valve
from opening when the reactor coolant system is at high pressure
conditions. Premature opening of the emergency core cooling system main
valves could result in fuel damage. The IAB valve then opens at reduced
reactor coolant system pressure to allow the main valve to open and
permit natural circulation cooling of the reactor core in response to a
plant event. Although the valve assemblies are considered an active
component, NuScale Power does not apply the single failure criterion to
the IAB valve, including to the IAB valve's function to close.
Consistent with Commission safety goals and the practice of risk-
informed decisionmaking, the NRC evaluated the NuScale emergency core
cooling system valve system without assuming a single active failure of
the IAB valve to close.
During design demonstration tests of the first-of-a-kind emergency
core cooling system valve system performed under Sec. 50.43(e),
NuScale Power implemented design modifications to the main valve and
IAB valve to demonstrate that the IAB valve will operate within a
specific design pressure range. The DCD specifies that the emergency
core cooling system valves (including the IAB valves) will be qualified
under American Society of Mechanical Engineers Standard QME-1-2007,
``Qualification of Active Mechanical Equipment Used in Nuclear Power
Plants,'' as endorsed by NRC Regulatory Guide 1.100, Revision 3,
``Seismic Qualification of Electrical and Active Mechanical Equipment
and Functional Qualification of Active Mechanical Equipment for Nuclear
Power Plants,'' prior to installation in a NuScale nuclear power plant.
Additionally, the NRC regulations in Sec. 50.55a require that a
NuScale nuclear power plant meet the requirements of the American
Society of Mechanical Engineers Operation and Maintenance of Nuclear
Power Plants, Division 1, OM Code: Section IST (OM Code) as
incorporated by reference in Sec. 50.55a for inservice testing of the
emergency core cooling system valves, unless relief is granted or an
alternative is authorized by the NRC. The NRC's review and findings
related to the IAB valve are documented in safety evaluation report
Chapter 3, ``Design of Structures, Systems, Components and Equipment,''
Section 3.9.6, ``Functional Design, Qualification, and Inservice
Testing Programs for Pumps, Valves, and Dynamic Restraints.'' These
findings show that the NRC regulatory requirements and DCD Part 2, Tier
2 provisions provide reasonable assurance that the emergency core
system valve system will be capable of performing its design-basis
functions in light of the safety significance of the required opening
and closing pressures for the individual IAB valves.
Further, Chapter 15, ``Transient and Accident Analyses,'' Section
15.0.0.5, ``Limiting Single Failures,'' of the safety evaluation report
states that the IAB valve is a first-of-a-kind, safety-significant,
active component integral to the NuScale emergency core cooling system.
NuScale Power does not apply the single failure criterion to the IAB
valve, and, on July 2, 2019, the Commission directed the staff in SRM-
SECY-19-0036, ``Staff Requirements--SECY-19-0036--Application of the
Single Failure Criterion to NuScale Power LLC's Inadvertent Actuation
Block Valves,'' to ``review Chapter 15 of the NuScale Design
Certification Application without assuming a single active failure of
the inadvertent actuation block valve to close.'' The Commission
further stated that ``[t]his approach is consistent with the
Commission's safety goal policy and associated core damage and large
release frequency goals and existing Commission direction on the use of
risk-informed decision-making, as articulated in the 1995 Policy
Statement on the Use of Probabilistic Risk Assessment Methods in
Nuclear Regulatory Activities and the White Paper on Risk-Informed and
Performance-Based Regulation (in SRM-SECY-98-144, ``White Paper on
Risk-Informed and Performance-Based Regulation,'' and Yellow
Announcement 99-019).''
Based on the NRC's historic application of the single failure
criterion and Commission direction on the subject, as described in
SECY-77-439, ``Single Failure Criterion''; SRM-SECY-94-084, ``Policy
and Technical Issues associated with the Regulatory Treatment of Non-
Safety Systems and Implementation of Design Certification and Light-
Water Reactor Design Issues''; and SRM-SECY-19-0036, the NRC has
retained discretion, in fact or application-specific circumstances, to
decide when to apply the single failure criterion. The Commission's
decision in SRM-SECY-19-0036 provides direction regarding the
appropriate application and interpretation of the regulatory
requirements in 10 CFR part 50 to the NuScale IAB valve's function to
close. This decision is similar to those in
[[Page 3293]]
previous Commission documents that addressed the use of the single
failure criterion and provided clarification on when to apply the
single failure criterion in other specific instances.
D. Conformance With General Design Criterion 27, ``Combined Reactivity
Control Systems Capability''
NuScale Power determined that, under certain end-of-cycle scenarios
with one control rod stuck out, the NuScale reactivity control systems
could not prevent re-criticality and return to power. This result does
not meet GDC 27 of appendix A to 10 CFR part 50, which covers
reactivity control systems to reliably control reactivity changes under
postulated accident conditions with margin for stuck control rods.
Therefore, NuScale Power submitted an exemption request for GDC 27
(refer to Section 15, ``10 CFR 50, Appendix A, Criterion 27, `Combined
Reactivity Control Systems Capability,' '' of DCA Part 7,
``Exemptions'').
NuScale Power analyses determined that the specified acceptable
fuel design limits would not be exceeded and that core cooling would be
maintained during a return to power under these scenarios. The global
core power level would be less than 10 percent and within capacity of
the safety-related, passive decay heat removal system. The NRC
independently verified NuScale Power's results and found that NuScale
achieves the fundamental safety functions for nuclear reactor safety,
which are to control heat generation, remove heat, and limit the
release of radioactive materials. Chapter 15, Section 15.0.6.4.1, of
the safety evaluation report contains details of the evaluation of this
exemption request. Additional information is provided in SECY-18-0099,
``NuScale Power Exemption Request from 10 CFR part 50, Appendix A,
General Design Criterion 27, `Combined Reactivity Control Systems
Capability,' '' dated October 9, 2018. The NRC granted the exemption
request.
E. Absence of Safety-Related Class 1E AC or DC Electrical Power
NuScale does not contain safety-related Class 1E AC or DC
electrical power systems. The purpose of appendix A to 10 CFR part 50,
GDC 17, ``Electric Power Systems,'' is to ensure that sufficient
electric power is available to accomplish plant functions important to
safety. NuScale provides passive safety systems and features to
accomplish plant safety-related functions without reliance on
electrical power.
NuScale incorporates several innovative features that reduce the
overall complexity of the design and lower the number of safety-related
systems necessary to mitigate postulated accidents. NuScale has no
safety-related functions that rely on electrical power. For example,
the emergency core cooling system performs its safety function without
reliance on safety-related electrical power or external sources of
coolant inventory makeup. NuScale Power provided a methodology to
substantiate its assertion that the safety-related systems do not rely
on Class 1E electrical power in topical report TR-0815-16497, Revision
1, ``Safety Classification of Passive Nuclear Power Plant Electrical
Systems,'' dated February 7, 2017. The NRC reviewed topical report TR-
0815-16497 and concluded that NuScale Power demonstrated that the
safety-related systems do not rely on Class 1E electrical power. The
NRC's review and conclusions are documented in a safety evaluation
report approving topical report TR-0815-16497, issued December 13,
2017, as described in the final safety evaluation report for Chapter 1,
``Introduction and General Discussion,'' and included in the approved
version of the topical report, TR-0815-16497-NP-A.
Because no safety-related functions of NuScale rely on electrical
power, NuScale does not need any safety-related electrical power
systems. Therefore, NuScale Power requested an exemption from GDC 17,
which requires the provision of onsite and offsite power to provide
sufficient capacity and capability to assure that (1) specified
acceptable fuel design limits and design conditions of the reactor
coolant pressure boundary are not exceeded as a result of anticipated
operational occurrences and (2) the core is cooled and containment
integrity and other vital functions are maintained in the event of
postulated accidents. The NRC determined that, subject to limitations
and conditions stipulated in its safety evaluation report for TR-0815-
16497, the underlying purpose of GDC 17 (to ensure sufficient electric
power is available to accomplish the safety functions of the respective
systems), is met without reliance on Class 1E electric power. In other
words, the onsite and offsite electric power systems are classified as
non-Class 1E systems and electric power is not needed (1) to achieve or
maintain safe shutdown, (2) to assure specified acceptable fuel design
limits and design conditions of the reactor coolant pressure boundary
are not exceeded as a result of anticipated operational occurrences, or
(3) to maintain core cooling, containment integrity, and other vital
functions during postulated accidents. Further, the onsite and offsite
power systems are not needed to permit functioning of structures,
systems, and components important to safety. Therefore, NuScale Power
was granted an exemption from GDC 17. The NRC's evaluation of NuScale
Power's exemption request from the requirements of GDC 17 is documented
in Section 8.1.5, ``Technical Evaluation for Exemptions,'' of the final
safety evaluation report for Chapter 8, ``Electric Power.''
F. Accident Source Term Methodology
The NRC reviewed NuScale Power's methods for developing accident
source terms and performing accident radiological consequence analyses.
As defined in Sec. 50.2, ``Definitions,'' a source term ``refers to
the magnitude and mix of the radionuclides released from the fuel,
expressed as fractions of the fission product inventory in the fuel, as
well as their physical and chemical form, and the timing of their
release.'' NuScale Power developed source terms for deterministic
accidents for NuScale that are similar to those that have been used in
safety and siting assessments for large light water reactors. The
design-basis accidents for NuScale are the main steam line break
outside containment, rod ejection accident, fuel handling accident,
steam generator tube failure, and the failure of small lines carrying
primary coolant outside containment.
To address the source term regulatory requirements, NuScale Power
submitted topical report TR-0915-17565, Revision 3, ``Accident Source
Term Methodology,'' dated April 2019. The topical report proposes a
methodology to develop a source term based on several severe accident
scenarios that result in core damage, taken from the design
probabilistic risk assessment. This source term is the surrogate
radiological source term for a core damage event.
The topical report also provides methods for determining radiation
sources not developed from core damage scenarios for use in the
evaluation of environmental qualification of equipment under Sec.
50.49, ``Environmental qualification of electric equipment important to
safety for nuclear power plants.'' Specifically, the report describes
an iodine spike source term not involving core damage, which is a
surrogate accident that bounds potential accidents with release of the
reactor coolant into the containment vessel.
[[Page 3294]]
The NRC staff submitted a related information paper to the
Commission, SECY-19-0079, ``Staff Approach to Evaluate Accident Source
Terms for the NuScale Power Design Certification Application,'' dated
August 16, 2019, describing the regulatory and technical issues raised
by unique aspects of NuScale Power's methodology and the staff's
approach to reviewing topical report TR-0915-17565.
The NRC's review and findings of topical report TR-0915-17565,
Revision 3, are documented in the topical report final safety
evaluation report issued on October 24, 2019. The approved version of
topical report TR-0915-17565-NP-A, Revision 4, is discussed in the
final safety evaluation report Section 12.2, ``Radiation Sources,''
Section 12.3, ``Radiation Protection Design Features,'' Section 3.11
``Environmental Qualification of Mechanical and Electrical Equipment,''
Section 15.0.2, ``Review of Transient and Accident Analysis Methods,''
and Section 15.0.3, ``Radiological Consequences of Design Basis
Accidents.'' The NRC found the accident source terms acceptable for the
purposes described in each of the above safety evaluation report
sections.
G. Boron Redistribution During Passive Cooling Modes
The NRC evaluated the effects of boron volatility and
redistribution during long term passive cooling. During this mode of
operation, boron-free steam will enter the downcomer and containment,
which can potentially challenge reactor core shutdown margin and could
lead to a return to power. The NRC reviewed analyses provided by
NuScale Power demonstrating that the reactor remains subcritical and
that specified acceptable fuel design limits are not exceeded. The NRC
evaluated the technical basis for NuScale Power's approach and
conducted confirmatory calculations and independent assessments to
determine its acceptability. The staff's review is primarily documented
in Chapter 15, Section 15.0.5, ``Long Term Decay Heat and Residual Heat
Removal,'' and Section 15.6.5, ``Loss of Coolant Accidents Resulting
from Spectrum of Postulated Piping Breaks within the Reactor Coolant
Pressure Boundary,'' of the safety evaluation report. Specifically, the
staff concluded that the top of active fuel remains covered with
acceptably low cladding temperatures and that for beginning-of-cycle
and middle-of-cycle conditions, with no operator actions, the core
remains subcritical. The potential for an end-of-cycle return to power
is discussed in Section IV.D, ``Conformance with General Design
Criterion 27, `Combined Reactivity Control Systems Capability,' '' of
this document. In addition, Chapter 19, Section 19.1.4.6.4, ``Success
Criteria, Accident Sequences, and Systems Analyses,'' of the safety
evaluation report concludes that an operator error during recovery of
the module from an uneven boron distribution scenario is unlikely to
lead to core damage and is not a significant risk contributor.
H. Exemptions
NuScale Power submitted a total of 17 requests for exemptions from
the following regulations, including those discussed as part of the
significant technical issues mentioned previously (see Table 1.14-1,
``NuScale Design Certification Exemptions,'' in Chapter 1 of the final
safety evaluation report):
1. Sec. Sec. 50.46a and 50.34(f)(2)(vi) (Reactor Coolant System
Venting)
2. Sec. 50.44 (Combustible Gas Control)
3. Sec. 50.62(c)(1) (Reduction of Risk from Anticipated Transients
Without Scram)
4. Appendix A to 10 CFR part 50, GDC 17, ``Electric Power Systems'';
GDC 18, ``Inspection and Testing of Electric Power Systems''; and
related provisions of GDC 34, ``Residual Heat removal''; GDC 35,
``Emergency Core Cooling''; GDC 38, ``Containment Heat Removal''; GDC
41, ``Containment Atmosphere Cleanup''; and GDC 44, ``Cooling Water''
(Electric Power Systems GDCs)
5. Appendix A to 10 CFR part 50, GDC 33, ``Reactor Coolant Makeup''
6. Sec. 50.54(m) (Control Room Staffing) (Alternative to meet the
regulation)
7. Appendix A to 10 CFR part 50, GDC 52, ``Capability for Containment
Leakage Rate Testing'' and Appendix J to 10 CFR part 50 (Type A
testing)
8. Appendix A to 10 CFR part 50, GDC 40, ``Testing of Containment Heat
Removal System''
9. Appendix A to 10 CFR part 50, GDC 55, ``Reactor Coolant Pressure
Boundary Penetrating Containment,'' GDC 56, ``Primary Containment
Isolation,'' and GDC 57, ``Closed Systems Isolation Valves''
(Containment Isolation)
10. Appendix K to 10 CFR part 50 (Emergency Core Cooling System
Evaluation Models)
11. Sec. 50.34(f)(2)(xx) (Power Supplies for Pressurizer Relief
Valves, Block Valves, and Level Indicators)
12. Sec. 50.34(f)(2)(xiii) (Pressurizer Heater Power Supplies)
13. Sec. 50.34(f)(2)(xiv)(E) (Containment Evacuation System Isolation)
14. Sec. 50.46 (Fuel Rod Cladding Material)
15. Appendix A to 10 CFR part 50, GDC 27, ``Combined Reactivity Control
Systems Capability''
16. Sec. 50.34(f)(2)(viii) (Post-Accident Sampling)
17. Appendix A to 10 CFR part 50, GDC 19, ``Control Room''
NRC's safety evaluation report for Chapter 1, ``Introduction and
General Discussion,'' Section 1.14, ``Index of Exemptions,'' lists
these exemption requests with the corresponding sections of the safety
evaluation report where these exemption requests have been evaluated.
The NRC granted each exemption request.
I. Differing Professional Opinion Related to Chapter 3 of NuScale
On September 17, 2020, a Differing Professional Opinion (DPO) was
submitted that raised concerns related to the seismic margin evaluation
of the NuScale reactor building and its structural response during the
review level earthquake. An ad-hoc review panel was formed and tasked
to review the DPO. The review panel subsequently issued its report to
the Director of the Office of Nuclear Reactor Regulation (NRR) on April
19, 2021. On May 19, 2021, the Director of NRR issued a decision to the
DPO submitter. For the reasons described in the decision, the Director
of NRR agreed with the review panel's finding that the NuScale reactor
building design was complete and acceptable for the purposes of a
design certification application. On June 14, 2021, the DPO submitter
appealed the DPO decision to the Executive Director for Operations
(EDO).
After consideration of the issues raised in the appeal, the EDO
issued a decision on the DPO appeal on February 8, 2022. The EDO
directed NRR to (1) document its evaluation of the stress averaging
approach used in the NuScale design certification application,
including, if necessary, updating the Final Safety Evaluation Report
and assess whether there are any impacts to the standard design
approval, and (2) evaluate and update guidance, or create knowledge
management tools, on how to assess applications that use stress
averaging for structural building design. On February 14, 2022, the DPO
submitter responded to the EDO's DPO appeal decision. In this response,
the submitter thanked the EDO for thoughtful consideration of the
concerns raised and provided clarification regarding the applicability
of the Probabilistic Risk Assessment-based seismic margin analysis to
the reactor building. After reviewing and considering the submitter's
response to
[[Page 3295]]
the DPO appeal decision, on March 15, 2022, the EDO directed the NRC
staff to review and consider the totality of the information provided
by the submitter when addressing the tasks mandated in the DPO appeal
decision.
In response to the EDO tasking, on May 13, 2022, the Director of
NRR issued a memo to the EDO (``Response to DPO Tasking'') discussing
the staff's review of the items described in the tasking, documenting
the staff's evaluation of the approach used in the NuScale design
certification, and detailing the staff's assessment of existing related
structural analysis guidance (ADAMS Accession No. ML22062A007). The
Director of NRR concluded that the staff sufficiently assessed the
evaluation of the demand (force/moment) averaging approach used in the
NuScale DCA; justified the acceptability to conclude that there are no
impacts to the NuScale standard design approval issued in September
2020; determined that an update or supplement to the final safety
evaluation report for the NuScale DCA is not necessary; and found that
the existing review guidance is sufficient to review and evaluate an
applicant's structural analysis/design. Details on the EDO's decision
on the DPO appeal and related correspondence, and the Response to DPO
Tasking are found in the information package for DPO-2020-004 (ADAMS
Accession No. ML22122A116).
The NRC staff's assessment of NuScale's use of the demand (force/
moment) averaging approach is documented in the Response to DPO
Tasking. The Response to DPO Tasking elaborates on the reasons for, but
does not change, the conclusion in the final safety evaluation report.
Based on this assessment, the NRC concludes that the use of the demand
(force/moment) averaging approach is acceptable, as stated in the final
safety evaluation report.
V. Discussion
Final Safety Evaluation Report
NuScale Power submitted the final revision of the NuScale DCA,
Revision 5, in July 2020 (ADAMS Accession No. ML20225A071). In August
2020, the NRC issued a final safety evaluation report after the
Advisory Committee on Reactor Safeguards (ACRS) performed its final
independent review and issued its July 29, 2020, letter to the
Commission on its findings and recommendations. The final safety
evaluation report is a collection of reports written by the NRC
documenting the safety findings from its review of the standard design
application, and it reflects all changes resulting from interactions
with the ACRS as well as changes in the final version of the DCA. The
final safety evaluation report, as elaborated on by the Response to DPO
Tasking, reflects that NuScale Power has resolved all technical and
safety issues with the exception of the three issues discussed
previously. As noted above, the Response to DPO Tasking elaborates on
the reasons for, but does not change, the conclusion in the final
safety evaluation report that NuScale's use of the demand (force/
moment) averaging approach is acceptable as a realistic engineering
practice.
In addition, the final safety evaluation report describes the
portions of the design that are not receiving finality in this rule
and, therefore, are not part of the certified design. The final safety
evaluation report also includes an index of all NRC requests for
additional information, a chronology of all documents related to the
NuScale DCA review, and summaries of public meetings and audits.
NuScale Design Certification Final Rule
This section describes the purpose and key aspects of each section
of this NuScale design certification final rule. All section and
paragraph references are to the provisions being added as appendix G to
10 CFR part 52, unless otherwise noted. The NRC has modeled this
NuScale design certification final rule on existing design
certification rules, with certain modifications where necessary to
account for differences in the design documentation, design features,
and environmental assessment (including severe accident mitigation
design alternatives). As a result, design certification rules are
standardized to the extent practical.
A. Introduction (Section I)
The purpose of Section I of appendix G to 10 CFR part 52 is to
identify the standard design that is approved by this design
certification final rule and the applicant for certification of the
standard design. Identification of the design certification applicant
is necessary to implement appendix G to 10 CFR part 52 for two reasons.
First, the implementation of Sec. 52.63(c) depends on whether an
applicant contracts with the design certification applicant to obtain
the generic DCD and supporting design information. If a COL applicant
does not use the design certification applicant to provide the design
information and instead uses an alternate vendor, then the COL
applicant must meet the requirements in Sec. 52.73. Second, paragraph
X.A.1 requires that the identified design certification applicant
maintain the generic DCD throughout the time that appendix G to 10 CFR
part 52 may be referenced.
B. Definitions (Section II)
The purpose of Section II of appendix G to 10 CFR part 52 is to
define specific terminology with respect to this design certification
final rule. During development of the first two design certification
rules, the NRC decided that there would be both generic DCDs maintained
by the NRC and the design certification applicant, as well as
individual plant-specific DCDs maintained by each applicant or licensee
that references a 10 CFR part 52 appendix. This distinction is
necessary in order to specify the relevant plant-specific requirements
to applicants and licensees referencing appendix G to 10 CFR part 52.
In order to facilitate the maintenance of the generic DCDs, the NRC
requires that applicants for a standard design certification update
their application to include an electronic copy of the final version of
the DCD. The final version incorporates all amendments to the DCA
submitted since the original application and any changes directed by
the NRC as a result of its review of the original DCA or as a result of
public comments. This final version is then incorporated by reference
in the design certification rule. Once incorporated by reference, the
final version becomes the ``generic DCD,'' which will be maintained by
the design certification applicant and the NRC and updated as needed to
include any generic changes made after this design certification
rulemaking. These changes would occur as the result of generic
rulemaking by the NRC, under the change criteria in Section VIII of
appendix G to 10 CFR part 52.
The NRC also requires each applicant and licensee referencing
appendix G to 10 CFR part 52 to submit and maintain a plant-specific
DCD as part of the COL final safety analysis report. The plant-specific
DCD must either include or incorporate by reference the information in
the generic DCD. The COL licensee is required to maintain the plant-
specific DCD, updating it as necessary to reflect the generic changes
to the DCD that the NRC may adopt through rulemaking, plant-specific
departures from the generic DCD that the NRC imposes on the licensee by
order, and any plant-specific departures that the licensee chooses to
make in accordance with the relevant processes in Section VIII of
appendix G to 10 CFR part 52. A COL applicant will also have to include
considerations for a multi-unit site in
[[Page 3296]]
the plant-specific DCD that were not previously evaluated as part of
the design certification rule, e.g., construction impacts on operating
units. Therefore, the plant-specific DCD functions like an updated
final safety analysis report because it would provide the most complete
and accurate information on a plant's design basis for that part of the
plant that would be within the scope of appendix G to 10 CFR part 52.
The NRC is treating the technical specifications in Part 4,
``Technical Specifications,'' of the DCA as a special category of
information and designating them as generic technical specifications in
order to facilitate the special treatment of this information under
appendix G to 10 CFR part 52. A COL applicant must submit plant-
specific technical specifications that consist of the generic technical
specifications, which may be modified as specified in paragraph VIII.C,
and the remaining site-specific information needed to complete the
technical specifications. The final safety analysis report that is
required by Sec. 52.79 will consist of the plant-specific DCD, the
site-specific final safety analysis report, and the plant-specific
technical specifications.
The terms Tier 1, Tier 2, and COL items (license information) are
defined in appendix G to 10 CFR part 52 because these concepts were not
envisioned when 10 CFR part 52 was developed. The design certification
applicants and the NRC use these terms in implementing a two-tiered
rule structure (the DCD is divided into Tier 1 and Tier 2 to support
the rule structure) that was proposed by representatives of the nuclear
industry after publication of 10 CFR part 52. The Commission approved
the use of the two-tiered rule structure in its staff requirements
memorandum (SRM), dated February 15, 1991, on SRM-SECY-90-377,
``Requirements for Design Certification under 10 CFR part 52,'' dated
November 8, 1990.
Tier 1 information means the portion of the design-related
information contained in the generic DCD that is approved and certified
by this appendix. Tier 2 information means the portion of the design-
related information contained in the generic DCD that is approved but
not certified by this appendix. The change process for Tier 2
information is similar, but not identical to, the change process set
forth in Sec. 50.59. The regulations in Sec. 50.59 describe when a
licensee may make changes to a plant as described in its final safety
analysis report without a license amendment. Because of some
differences in how the change control requirements are structured in
the design certification rules, certain definitions contained in Sec.
50.59 are not applicable to 10 CFR part 52 and are not being included
in this final rule. The NRC is including a definition for ``Departure
from a method of evaluation'' in paragraph II.F of appendix G to 10 CFR
part 52, so that the eight criteria in paragraph VIII.B.5.b will be
implemented for new reactors as intended.
C. Scope and Contents (Section III)
The purpose of Section III of appendix G to 10 CFR part 52 is to
describe and define the scope and content of this design certification,
explain how to obtain a copy of the generic DCD, identify requirements
for incorporation by reference of the design certification rule, and
set forth how documentation discrepancies or inconsistencies are to be
resolved.
Paragraph III.A is the required statement of the Office of the
Federal Register for approval of the incorporation by reference of the
NuScale DCD, Revision 5. In addition, this paragraph provides the
information on how to obtain a copy of the DCD. Unlike previous design
certifications, the documents submitted to the NRC by NuScale Power did
not use the title ``Design Control Document;'' they used the title
``Design Certification Application'' instead.
Paragraph III.B is the requirement for applicants and licensees
referencing appendix G to 10 CFR part 52. The legal effect of
incorporation by reference is that the incorporated material has the
same legal status as if it were published in the Code of Federal
Regulations. This material, like any other properly issued regulation,
has the force and effect of law. Tier 1 and Tier 2 information
(including the technical and topical reports referenced in the DCD Tier
2, Chapter 1) and generic technical specifications have been combined
into a single document called the generic DCD in order to effectively
control this information and facilitate its incorporation by reference
into the rule. In addition, paragraph III.B clarifies that the
conceptual design information and NuScale Power's evaluation of severe
accident mitigation design alternatives are not considered to be part
of appendix G to 10 CFR part 52. As provided by Sec. 52.47(a)(24),
these conceptual designs are not part of appendix G to 10 CFR part 52
and, therefore, are not applicable to an application that references
appendix G to 10 CFR part 52. Therefore, an applicant would not be
required to conform to the conceptual design information that was
provided by the design certification applicant. The conceptual design
information, which consists of site-specific design features, was
required to facilitate the design certification review. Similarly, the
severe accident mitigation design alternatives were required to
facilitate the environmental assessment.
Paragraphs III.C and III.D set forth the manner by which potential
conflicts are to be resolved and identify the controlling document.
Paragraph III.C establishes the Tier 1 description in the DCD as
controlling in the event of an inconsistency between the Tier 1 and
Tier 2 information in the DCD. Paragraph III.D establishes the generic
DCD as the controlling document in the event of an inconsistency
between the DCD and the final safety evaluation report for the
certified standard design.
Paragraph III.E makes it clear that design activities outside the
scope of the design certification may be performed using actual site
characteristics. This provision applies to site-specific portions of
the plant, such as the administration building.
D. Additional Requirements and Restrictions (Section IV)
Section IV of appendix G to 10 CFR part 52 sets forth additional
requirements and restrictions imposed upon an applicant who references
appendix G to 10 CFR part 52.
Paragraph IV.A sets forth the information requirements for COL
applicants and distinguishes between information and documents that
must be included in the application or the DCD and those which may be
incorporated by reference. Any incorporation by reference in the
application should be clear and should specify the title, date, edition
or version of a document, the page number(s), and table(s) containing
the relevant information to be incorporated. The legal effect of such
an incorporation by reference into the application is that appendix G
to 10 CFR part 52 would be legally binding on the applicant or
licensee.
In paragraph IV.B the NRC reserves the right to determine how
appendix G to 10 CFR part 52 may be referenced under 10 CFR part 50.
This determination may occur in the context of a subsequent rulemaking
modifying 10 CFR part 52 or this design certification rule, or on a
case-by-case basis in the context of a specific application for a 10
CFR part 50 construction permit or operating license. This provision is
necessary because the previous design certification rules were not
[[Page 3297]]
implemented in the manner that was originally envisioned at the time
that 10 CFR part 52 was issued. The NRC's concern is with the manner by
which the inspections, tests, analyses, and acceptance criteria (ITAAC)
were developed and the lack of experience with design certifications in
a licensing proceeding. Therefore, it is appropriate that the NRC
retain some discretion regarding the manner by which appendix G to 10
CFR part 52 could be referenced in a 10 CFR part 50 licensing
proceeding.
In paragraph IV.C, the NRC lists design-specific regulations that
apply to licenses that reference this appendix.
E. Applicable Regulations (Section V)
The purpose of Section V of appendix G to 10 CFR part 52 is to
specify the regulations that were applicable and in effect at the time
this design certification was approved. These regulations consist of
the technically relevant regulations identified in paragraph V.A,
except for the regulations in paragraph V.B that would not be
applicable to this certified design.
F. Issue Resolution (Section VI)
The purpose of Section VI of appendix G to 10 CFR part 52 is to
identify the scope of issues that are resolved by the NRC through this
final rule and, therefore, are ``matters resolved'' within the meaning
and intent of Sec. 52.63(a)(5). The section is divided into five
parts: paragraph VI.A identifies the NRC's safety findings in adopting
appendix G to 10 CFR part 52, paragraph VI.B identifies the scope and
nature of issues that are resolved by this final rule, paragraph VI.C
identifies issues that are not resolved by this final rule, and
paragraph VI.D identifies the issue finality restrictions applicable to
the NRC with respect to appendix G to 10 CFR part 52.
Paragraph VI.A describes the nature of the NRC's findings in
general terms and makes the findings required by Sec. 52.54 for the
NRC's approval of this design certification final rule.
Paragraph VI.B sets forth the scope of issues that may not be
challenged as a matter of right in subsequent proceedings. The
introductory phrase of paragraph VI.B clarifies that issue resolution,
as described in the remainder of the paragraph, extends to the
delineated NRC proceedings referencing appendix G to 10 CFR part 52.
The remainder of paragraph VI.B describes the categories of information
for which there is issue resolution.
Paragraph VI.C reserves the right of the NRC to impose operational
requirements on applicants that reference appendix G to 10 CFR part 52.
This provision reflects the fact that only some operational
requirements, including portions of the generic technical specification
in Chapter 16 of the DCD, were completely or comprehensively reviewed
by the NRC in this design certification final rule proceeding. The NRC
notes that operational requirements may be imposed on licensees
referencing this design certification through the inclusion of license
conditions in the license or inclusion of a description of the
operational requirement in the plant-specific final safety analysis
report.\1\ The NRC's choice of the regulatory vehicle for imposing the
operational requirements will depend upon, among other things, (1)
whether the development and/or implementation of these requirements
must occur prior to either the issuance of the COL or the Commission
finding under Sec. 52.103(g), and (2) the nature of the change
controls that are appropriate given the regulatory, safety, and
security significance of each operational requirement.
---------------------------------------------------------------------------
\1\ Certain activities ordinarily conducted following fuel load
and, therefore, considered ``operational requirements,'' but which
may be relied upon to support a Commission finding under Sec.
52.103(g), may themselves be the subject of ITAAC to ensure their
implementation prior to the Sec. 52.103(g) finding.
---------------------------------------------------------------------------
Also, paragraph VI.C allows the NRC to impose future operational
requirements (distinct from design matters) on applicants who reference
this design certification. License conditions for portions of the plant
within the scope of this design certification (e.g., startup and power
ascension testing) are not restricted by Sec. 52.63. The requirement
to perform these testing programs is contained in the Tier 1
information. However, ITAAC cannot be specified for these subjects
because the matters to be addressed in these license conditions cannot
be verified prior to fuel load and operation when the ITAAC are
satisfied. In the absence of detailed design information to evaluate
the need for and develop specific post-fuel load verifications for
these matters, the NRC is reserving the right to impose, at the time of
COL issuance, license conditions addressing post-fuel load verification
activities for portions of the plant within the scope of this design
certification.
Paragraph VI.D reiterates the restrictions (contained in Section
VIII of appendix G to 10 CFR part 52) placed upon the NRC when ordering
generic or plant-specific modifications, changes, or additions to
structures, systems, and components, design features, design criteria,
and ITAAC within the scope of the certified design.
Paragraph VI.E provides that the NRC will specify at an appropriate
time the procedures on how to obtain access to sensitive unclassified
and non-safeguards information (SUNSI) and safeguards information (SGI)
for the NuScale design certification rule. Access to such information
would be for the sole purpose of requesting or participating in certain
specified hearings, such as hearings required by Sec. 52.85 or an
adjudicatory hearing. For proceedings where the notice of hearing was
published before the effective date of the final rule, the Commission's
order governing access to SUNSI and SGI shall be used to govern access
to such information within the scope of the rulemaking. For proceedings
in which the notice of hearing or opportunity for hearing is published
after the effective date of the final rule, paragraph VI.E applies and
governs access to SUNSI and SGI.
G. Duration of This Appendix (Section VII)
The purpose of Section VII of appendix G to 10 CFR part 52 is, in
part, to specify the period during which this design certification may
be referenced by an applicant, under Sec. 52.55, and the period it
will remain valid when the design certification is referenced. For
example, if an application references this design certification during
the 15-year period, then the design certification would be effective
until the application is withdrawn or the license issued on that
application expires. The NRC intends for appendix G to 10 CFR part 52
to remain valid for the life of any license that references the design
certification to achieve the benefits of standardization and licensing
stability. This means that changes to, or plant-specific departures
from, information in the plant-specific DCD must be made under the
change processes in Section VIII for the life of the plant.
H. Processes for Changes and Departures (Section VIII)
The purpose of Section VIII of appendix G to 10 CFR part 52 is to
set forth the processes for generic changes to, or plant-specific
departures (including exemptions) from, the DCD. The NRC adopted this
restrictive change process in order to achieve a more stable licensing
process for applicants and licensees that reference design
certification rules. Section VIII is divided into three paragraphs,
which correspond to Tier 1, Tier 2, and operational requirements.
[[Page 3298]]
Generic changes (called ``modifications'' in Sec. 52.63(a)(3))
must be accomplished by rulemaking because the intended subject of the
change is this design certification rule itself, as is contemplated by
Sec. 52.63(a)(1). Consistent with Sec. 52.63(a)(3), any generic
rulemaking changes are applicable to all plants, absent circumstances
which render the change technically irrelevant. By contrast, plant-
specific departures could be required by either an order to one or more
applicants or licensees; or an applicant or licensee-initiated
departure applicable only to that applicant's or licensee's plant(s),
similar to a Sec. 50.59 departure or an exemption. Because these
plant-specific departures will result in a DCD that is unique for that
plant, Section X requires an applicant or licensee to maintain a plant-
specific DCD. For purposes of brevity, the following discussion refers
to the processes for both generic changes and plant-specific departures
as ``change processes.'' Section VIII refers to an exemption from one
or more requirements of this appendix and addresses the criteria for
granting an exemption. The NRC cautions that when the exemption
involves an underlying substantive requirement (i.e., a requirement
outside this appendix), then the applicant or licensee requesting the
exemption must demonstrate that an exemption from the underlying
applicable requirement meets the criteria of Sec. Sec. 52.7 and 50.12.
For the NuScale review, the staff followed the approach described
in SECY-17-0075, ``Planned Improvements in Design Certification Tiered
Information Designations,'' dated July 24, 2017, to evaluate the
applicant's designation of information as Tier 1 or Tier 2 information.
Unlike some of the prior DCAs, this application did not contain any
Tier 2* information. As described in SECY-17-0075, prior design
certification rules in 10 CFR part 52, appendices A through E,
information contained in the DCD was divided into three designations:
Tier 1, Tier 2, and Tier 2*. Tier 1 information is the portion of
design-related information in the generic DCD that the Commission
approves in the 10 CFR part 52 design certification rule appendices. To
change Tier 1 information, NRC approval by rulemaking or approval of an
exemption from the certified design rule is required. Tier 2
information is also approved by the Commission in the 10 CFR part 52
design certification rule appendices, but it is not certified and
licensees who reference the design can change this information using
the process outlined in Section VIII of the appendices. This change
process is similar to that in Sec. 50.59 and is generally referred to
as the ``50.59-like'' process. If the criteria in Section VIII are met,
a licensee can change Tier 2 information without prior NRC approval.
As mentioned in the previous paragraph, the NRC created a third
category, Tier 2*, in other design certification rules. This third
category was created to address industry requests to minimize the scope
of Tier 1 information and provide greater flexibility for making
changes. Unlike Tier 2 information, all changes to Tier 2* information
require a license amendment, but unlike Tier 1 information, no
exemption is required. In those rules, Tier 2* information has the same
safety significance as Tier 1 information but is part of the Tier 2
section of the DCD to afford more flexibility for licensees to change
this type of information.
The applicant did not designate or categorize any Tier 2*
information in the NuScale DCA. The NRC evaluated the Tier 2
information to determine whether any of that information should require
NRC approval before it is changed. If the NRC had identified any such
information in Tier 2, then the NRC would have requested that the
applicant revise the application to categorize that information as Tier
1 or Tier 2*. The NRC did not identify any information in Tier 2 that
should be categorized as Tier 2*. Because neither the applicant nor the
NRC have designated any information in the DCD as Tier 2*, that
designation and related requirements are not being used in this design
certification rule.
Tier 1 Information
Paragraph A of Section VIII describes the change process for
changes to Tier 1 information that are accomplished by rulemakings that
amend the generic DCD and are governed by the standards in Sec.
52.63(a)(1). A generic change under Sec. 52.63(a)(1) will not be made
to a certified design while it is in effect unless the change: (1) is
necessary for compliance with NRC regulations applicable and in effect
at the time the certification was issued; (2) is necessary to provide
adequate protection of the public health and safety or common defense
and security; (3) reduces unnecessary regulatory burden and maintains
protection to public health and safety and common defense and security;
(4) provides the detailed design information necessary to resolve
select design acceptance criteria; (5) corrects material errors in the
certification information; (6) substantially increases overall safety,
reliability, or security of a facility and the costs of the change are
justified; or (7) contributes to increased standardization of the
certification information. The rulemakings must provide for notice and
opportunity for public comment on the proposed change under Sec.
52.63(a)(2). The NRC will give consideration as to whether the benefits
justify the costs for plants that are already licensed or for which an
application for a permit or license is under consideration.
Departures from Tier 1 may occur in two ways: (1) the NRC may order
a licensee to depart from Tier 1, as provided in paragraph VIII.A.3; or
(2) an applicant or licensee may request an exemption from Tier 1, as
addressed in paragraph VIII.A.4. If the NRC seeks to order a licensee
to depart from Tier 1, paragraph VIII.A.3 would require that the NRC
find both that the departure is necessary for adequate protection or
for compliance and that special circumstances are present. Paragraph
VIII.A.4 provides that exemptions from Tier 1 requested by an applicant
or licensee are governed by the requirements of Sec. Sec. 52.63(b)(1)
and 52.98(f), which provide an opportunity for a hearing. In addition,
the NRC would not grant requests for exemptions that will result in a
significant decrease in the level of safety otherwise provided by the
design.
Tier 2 Information
Paragraph B of Section VIII describes the change processes for the
Tier 2 information, which have the same elements as the Tier 1 change
process, but some of the standards for plant-specific orders and
exemptions would be different. Generic Tier 2 changes would be
accomplished by rulemaking that would amend the generic DCD and would
be governed by the standards in Sec. 52.63(a)(1). A generic change
under Sec. 52.63(a)(1) would not be made to a certified design while
it is in effect unless the change: (1) is necessary for compliance with
NRC regulations that were applicable and in effect at the time the
certification was issued; (2) is necessary to provide adequate
protection of the public health and safety or common defense and
security; (3) reduces unnecessary regulatory burden and maintains
protection to public health and safety and common defense and security;
(4) provides the detailed design information necessary to resolve
select design acceptance criteria; (5) corrects material errors in the
certification information; (6)
[[Page 3299]]
substantially increases overall safety, reliability, or security of a
facility and the costs of the change are justified; or (7) contributes
to increased standardization of the certification information.
Departures from Tier 2 would occur in four ways: (1) the NRC may
order a plant-specific departure, as set forth in paragraph VIII.B.3;
(2) an applicant or licensee may request an exemption from a Tier 2
requirement as set forth in paragraph VIII.B.4; (3) a licensee may make
a departure without prior NRC approval under paragraph VIII.B.5; or (4)
the licensee may request NRC approval for proposed departures that do
not meet the requirements in paragraph VIII.B.5 as provided in
paragraph VIII.B.5.e.
Similar to ordered Tier 1 departures and generic Tier 2 changes,
ordered Tier 2 departures could not be imposed except when necessary,
either to bring the certification into compliance with the NRC's
regulations applicable and in effect at the time of approval of the
design certification or to ensure adequate protection of the public
health and safety or common defense and security, as set forth in
paragraph VIII.B.3. However, unlike Tier 1 departures, the Commission
would not have to consider whether the special circumstances for the
Tier 2 departures would outweigh any decrease in safety that may result
from the reduction in standardization caused by the plant-specific
order, as required by Sec. 52.63(a)(4). The NRC has determined that it
is not necessary to impose an additional limitation for standardization
similar to that imposed on Tier 1 departures by Sec. 52.63(a)(4) and
(b)(1) because it would unnecessarily restrict the flexibility of
applicants and licensees with respect to Tier 2 information.
An applicant or licensee may request an exemption from Tier 2
information as set forth in paragraph VIII.B.4. The applicant or
licensee would have to demonstrate that the exemption complies with one
of the special circumstances in regulations governing specific
exemptions in Sec. 50.12(a). In addition, the NRC would not grant
requests for exemptions that will result in a significant decrease in
the level of safety otherwise provided by the design. However, unlike
Tier 1 changes, the special circumstances for the exemption do not have
to outweigh any decrease in safety that may result from the reduction
in standardization caused by the exemption. If the exemption is
requested by an applicant for a license, the exemption would be subject
to litigation in the same manner as other issues in the licensing
hearing, consistent with Sec. 52.63(b)(1). If the exemption is
requested by a licensee, then the exemption would be subject to
litigation in the same manner as a license amendment.
Paragraph VIII.B.5 allows an applicant or licensee to depart from
Tier 2 information, without prior NRC approval, if it does not involve
a change to, or departure from, Tier 1 information, technical
specification, or does not require a license amendment under paragraphs
VIII.B.5.b or c. The technical specifications referred to in VIII.B.5.a
of this paragraph are the technical specifications in Chapter 16 of the
generic DCD, including bases, for departures made prior to the issuance
of the COL. After the issuance of the COL, the plant-specific technical
specifications would be controlling under paragraph VIII.B.5. The
requirement for a license amendment in paragraph VIII.B.5.b is similar
to the requirement in Sec. 50.59 and applies to all of the information
in Tier 2 except for the information that resolves the severe accident
issues or the information required by Sec. 52.47(a)(28) to address
aircraft impacts.
Paragraph VIII.B.5.d addresses information described in the DCD to
address aircraft impacts, in accordance with Sec. 52.47(a)(28). Under
Sec. 52.47(a)(28), applicants are required to include the information
required by Sec. 50.150(b) in their DCD. An applicant or licensee who
changes this information is required to consider the effect of the
changed design feature or functional capability on the original
aircraft impact assessment required by Sec. 50.150(a). The applicant
or licensee is also required to describe in the plant-specific DCD how
the modified design features and functional capabilities continue to
meet the assessment requirements in Sec. 50.150(a)(1). Submittal of
this updated information is governed by the reporting requirements in
Section X.B.
During an ongoing adjudicatory proceeding (e.g., for issuance of a
COL), a party who believes that an applicant or licensee has not
complied with paragraph VIII.B.5 when departing from Tier 2 information
may petition to admit such a contention into the proceeding under
paragraph VIII.B.5.g. As set forth in paragraph VIII.B.5.g, the
petition would have to comply with the NRC's hearing requirements at
Sec. 2.309 and show that the departure does not comply with paragraph
VIII.B.5. If on the basis of the petition and any responses thereto,
the presiding officer in the proceeding determines that the required
showing has been made, the matter would be certified to the Commission
for its final determination. In the absence of a proceeding, assertions
of nonconformance with paragraph VIII.B.5 requirements applicable to
Tier 2 departures would be treated as petitions for enforcement action
under Sec. 2.206.
Operational Requirements
The change process for technical specifications and other
operational requirements that were reviewed and approved in the design
certification rule is set forth in Section VIII, paragraph C. The key
to using the change processes described in Section VIII is to determine
if the proposed change or departure would require a change to a design
feature described in the generic DCD. If a design change is required,
then the appropriate change process in paragraph VIII.A or VIII.B would
apply. However, if a proposed change to the technical specifications or
other operational requirements does not require a change to a design
feature in the generic DCD, then paragraph VIII.C would apply. This
change process has elements similar to the Tier 1 and Tier 2 change
processes in paragraphs VIII.A and VIII.B, but with significantly
different change standards. Because of the different finality status
for technical specifications and other operational requirements, the
NRC designated a special category of information, consisting of the
technical specifications and other operational requirements, with its
own change process in paragraph VIII.C. The language in paragraph
VIII.C also distinguishes between generic (Chapter 16 of the DCD) and
plant-specific technical specifications to account for the different
treatment and finality consistent with technical specifications before
and after a license is issued.
The process in paragraph VIII.C.1 for making generic changes to the
generic technical specifications or other operational requirements in
the generic DCD is accomplished by rulemaking and governed by the
backfit standards in Sec. 50.109. The determination of whether the
generic technical specifications and other operational requirements
were completely reviewed and approved in the design certification rule
is based upon the extent to which the NRC reached a safety conclusion
in the final safety evaluation report on this matter. If a technical
specification or operational requirement was completely reviewed and
finalized in the design certification rule, then the requirement of
Sec. 50.109 would apply because a position was taken on that safety
matter. Generic changes made under paragraph
[[Page 3300]]
VIII.C.1 would be applicable to all applicants or licensees (refer to
paragraph VIII.C.2), unless the change is irrelevant because of a
plant-specific departure.
Some generic technical specifications contain values in brackets [
]. The brackets are placeholders indicating that the NRC's review is
not complete and represent a requirement that an applicant for a COL
referencing appendix G to 10 CFR part 52 must replace the values in
brackets with final plant-specific values (refer to guidance provided
in Regulatory Guide 1.206, Revision 1, ``Applications for Nuclear Power
Plants,'' dated October 2018). The values in brackets are neither part
of the design certification rule nor are they binding. Therefore, the
replacement of bracketed values with final plant-specific values does
not require an exemption from the generic technical specifications.
Plant-specific departures may occur by either an order under
paragraph VIII.C.3 or an applicant's exemption request under paragraph
VIII.C.4. The basis for determining if the technical specification or
operational requirement was completely reviewed and approved for these
processes would be the same as for paragraph VIII.C.1 previously
discussed. If the technical specification or operational requirement
was comprehensively reviewed and finalized in the design certification
rule, then the NRC must demonstrate that special circumstances are
present before ordering a plant-specific departure. If not, there would
be no restriction on plant-specific changes to the technical
specifications or operational requirements, prior to the issuance of a
license, provided a design change is not required. Although the generic
technical specifications were reviewed and approved by the NRC in
support of the design certification review, the NRC intends to consider
the lessons learned from subsequent operating experience during its
licensing review of the plant-specific technical specifications. The
process for petitioning to intervene on a technical specification or
operational requirement contained in paragraph VIII.C.5 is similar to
other issues in a licensing hearing, except that the petitioner must
also demonstrate why special circumstances are present pursuant to
Sec. 2.335.
Paragraph VIII.C.6 states that the generic technical specifications
would have no further effect on the plant-specific technical
specifications after the issuance of a license that references this
appendix and the change process. After a license is issued, the bases
for the plant-specific technical specification would be controlled by
the bases change provision set forth in the administrative controls
section of the plant-specific technical specifications.
I. [RESERVED] (Section IX)
This section is reserved for future use. The matters discussed in
this section of earlier design certification rules--inspections, tests,
analyses, and acceptance criteria--are now addressed in the substantive
provisions of 10 CFR part 52. Accordingly, there is no need to repeat
these regulatory provisions in the NuScale design certification rule.
However, this section is being reserved to maintain consistent section
numbering with other design certification rules.
J. Records and Reporting (Section X)
The purpose of Section X of appendix G to 10 CFR part 52 is to set
forth the requirements that will apply to maintaining records of
changes to and departures from the generic DCD, which are to be
reflected in the plant-specific DCD. Section X also sets forth the
requirements for submitting reports (including updates to the plant-
specific DCD) to the NRC. This section of appendix G to 10 CFR part 52
is similar to the requirements for records and reports in 10 CFR part
50, except for minor differences in information collection and
reporting requirements.
Paragraph X.A.1 requires that a generic DCD including referenced
SUNSI and SGI be maintained by the applicant for this final rule. The
generic DCD concept was developed, in part, to meet the requirements
for incorporation by reference, including public availability of
documents incorporated by reference. However, the SUNSI and SGI could
not be included in the generic DCD because they are not publicly
available. Nonetheless, the SUNSI and SGI were reviewed by the NRC and,
as stated in paragraph VI.B.2, the NRC would consider the information
to be resolved within the meaning of Sec. 52.63(a)(5). Because this
information, or its equivalent, is not in the generic DCD, it is
required to be provided by an applicant for a license referencing
appendix G to 10 CFR part 52. Only the generic DCD is identified and
incorporated by reference by this final rule. The generic DCD and the
NRC approved version of the SUNSI and SGI must be maintained by the
applicant (NuScale Power) for the period of time that appendix G to 10
CFR part 52 may be referenced.
Paragraphs X.A.2 and X.A.3 place recordkeeping requirements on the
applicant or licensee that reference this design certification so that
its plant-specific DCD accurately reflects both generic changes to the
generic DCD and plant-specific departures made under Section VIII. The
term ``plant-specific'' is used in paragraph X.A.2 and other sections
of appendix G to 10 CFR part 52 to distinguish between the generic DCD
that this final rule incorporates by reference into appendix G to 10
CFR part 52, and the plant-specific DCD that the COL applicant is
required to submit under paragraph IV.A. The requirement to maintain
changes to the generic DCD is explicitly stated to ensure that these
changes are not only reflected in the generic DCD, which will be
maintained by the applicant for the design certification, but also in
the plant-specific DCD. Therefore, records of generic changes to the
DCD will be required to be maintained by both entities to ensure that
both entities have up-to-date DCDs.
Paragraph X.A.4.a requires the design certification rule applicant
to maintain a copy of the aircraft impact assessment analysis for the
term of the certification and any renewal. This provision, which is
consistent with Sec. 50.150(c)(3), would facilitate any NRC
inspections of the assessment that the NRC decides to conduct.
Similarly, paragraph X.A.4.b requires an applicant or licensee who
references appendix G to 10 CFR part 52 to maintain a copy of the
aircraft impact assessment performed to comply with the requirements of
Sec. 50.150(a) throughout the pendency of the application and for the
term of the license and any renewal. This provision is consistent with
Sec. 50.150(c)(4). For all applicants and licensees, the supporting
documentation retained should describe the methodology used in
performing the assessment, including the identification of potential
design features and functional capabilities to show that the acceptance
criteria in Sec. 50.150(a)(1) will be met.
Paragraph X.A does not place recordkeeping requirements on site
specific information that is outside the scope of this rule. As
discussed in paragraph V.B of this document, the final safety analysis
report required by Sec. 52.79 will contain the plant-specific DCD and
the site-specific information for a facility that references this rule.
The phrase ``site specific portion of the final safety analysis
report'' in paragraph X.B.3.c refers to the information that is
contained in the final safety analysis report for a facility (required
by Sec. 52.79), but is not part of the plant-specific DCD (required by
paragraph IV.A). Therefore, this final rule does not require that
duplicate documentation be maintained by an applicant or licensee that
references this
[[Page 3301]]
rule because the plant-specific DCD is part of the final safety
analysis report for the facility.
Paragraph X.B.1 requires applicants or licensees that reference
this rule to submit reports that describe departures from the DCD and
include a summary of the written evaluations. The requirement for the
written evaluations is set forth in paragraph X.A.3. The frequency of
the report submittals is set forth in paragraph X.B.3. The requirement
for submitting a summary of the evaluations is similar to the
requirement in Sec. 50.59(d)(2).
Paragraph X.B.2 requires applicants or licensees that reference
this rule to submit updates to the DCD, which include both generic
changes and plant-specific departures, as set forth in paragraph X.B.3.
The requirements in paragraph X.B.3 for submitting reports will vary
according to certain time periods during a facility's lifetime. If a
potential applicant for a COL that references this rule decides to
depart from the generic DCD prior to submission of the application,
then paragraph X.B.3.a will require that the updated DCD be submitted
as part of the initial application for a license. Under paragraph
X.B.3.b, the applicant may submit any subsequent updates to its plant-
specific DCD along with its amendments to the application provided that
the submittals are made at least once per year.
Paragraph X.B.3.b also requires semi-annual submission of the
reports required by paragraphs X.B.1 and X.B.2 throughout the period of
application review and construction. The NRC will use the information
in the reports to support planning for the NRC's inspection and
oversight during this phase, when the licensee is conducting detailed
design, procurement of components and equipment, construction, and
preoperational testing. In addition, the NRC will use the information
in making its finding on ITAAC under Sec. 52.103(g), as well as any
finding on interim operation under Section 189.a(1)(B)(iii) of the
Atomic Energy Act of 1954, as amended. Once a facility begins operation
(for a COL under 10 CFR part 52, after the Commission has made a
finding under Sec. 52.103(g)), the frequency of reporting will be
governed by the requirements in paragraph X.B.3.c.
VI. Public Comment Analysis
The NRC prepared a summary and analysis of public comments received
on the 2021 proposed rule, as referenced in the ``Availability of
Documents'' section. The NRC received eight comment submissions during
the public comment period that ended on October 14, 2021, and one late-
filed comment submission on October 15, 2021, that the NRC was able to
include in its consideration for this final rule. A comment submission
is a communication or document submitted to the NRC by an individual or
entity, with one or more individual comments addressing a subject or
issue. Private citizens provided four comment submissions, nuclear
industry organizations provided two comment submissions, science
advocacy groups provided two comment submissions, and a labor union
provided one comment submission. Of the nine comments, six were in
favor of the design certification rule, one was opposed, and the other
two comment submittals posed questions but stated no preference for the
outcome of the rule. Six of the nine comment submissions contained
questions on technical aspects of the design, corrections to the
statement of considerations, and interpretation of requirements.
The public comment submittals are available on the Federal
rulemaking website under Docket ID NRC-2017-0029. NRC's response to the
public comments, including a summary of how NRC revised the proposed
rule in response to public input, can be found in the public comment
analysis document.
VII. Section-by-Section Analysis
The following paragraphs describe the specific changes in this
final rule: Section 52.11, Information collection requirements: Office
of Management and Budget (OMB) approval.
In Sec. 52.11, this final rule adds new appendix G to 10 CFR part
52 to the list of information collection requirements in paragraph (b)
of this section.
Appendix G to Part 52--Design Certification Rule for the NuScale
Standard Design
This final rule adds appendix G to 10 CFR part 52 to incorporate
the NuScale standard design into the NRC's regulations. Applicants
intending to construct and operate a plant using NuScale may do so by
referencing the design certification rule.
VIII. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC
certifies that this rule does not have a significant economic impact on
a substantial number of small entities. This final rule affects only
the licensing and operation of nuclear power plants. The companies that
own these plants do not fall within the scope of the definition of
``small entities'' set forth in the Regulatory Flexibility Act or the
size standards established by the NRC (Sec. 2.810).
IX. Regulatory Analysis
The NRC has not prepared a regulatory analysis for this final rule.
The NRC prepares regulatory analyses for rulemakings that establish
generic regulatory requirements applicable to all licensees. Design
certifications are not generic rulemakings in the sense that design
certifications do not establish standards or requirements with which
all licensees must comply. Rather, design certifications are NRC
approvals of specific nuclear power plant designs by rulemaking, which
then may be voluntarily referenced by applicants for combined licenses.
Furthermore, design certification rules are requested by an applicant
for a design certification, rather than the NRC. Preparation of a
regulatory analysis in this circumstance would not be useful because
the design to be certified is proposed by the applicant rather than the
NRC. For these reasons, the NRC concludes that preparation of a
regulatory analysis is neither required nor appropriate.
X. Backfitting and Issue Finality
The NRC has determined that this final rule does not constitute a
backfit as defined in the backfit rule (Sec. 50.109), and that it is
not inconsistent with any applicable issue finality provision in 10 CFR
part 52.
This initial design certification rule does not constitute
backfitting as defined in the backfit rule (Sec. 50.109) because there
are no operating licenses under 10 CFR part 50 referencing this design
certification final rule.
This initial design certification rule is not inconsistent with any
applicable issue finality provision in 10 CFR part 52 because it does
not impose new or changed requirements on existing design certification
rules in appendices A through F to 10 CFR part 52, and no combined
licenses, construction permits, or manufacturing licenses issued by the
NRC at this time reference this design certification final rule.
For these reasons, neither a backfit analysis nor a discussion
addressing the issue finality provisions in 10 CFR part 52 was prepared
for this final rule.
XI. Plain Writing
The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal
agencies to write documents in a clear, concise, well-organized manner
that also follows other best practices appropriate to the subject or
field and the intended audience. The NRC has written this
[[Page 3302]]
document to be consistent with the Plain Writing Act as well as the
Presidential Memorandum, ``Plain Language in Government Writing,''
published June 10, 1998 (63 FR 31883).
XII. Environmental Assessment and Finding of No Significant Impact
The NRC conducted an environmental assessment and has determined
under the National Environmental Policy Act of 1969, as amended (NEPA),
and the NRC's regulations in subpart A of 10 CFR part 51, that this
final rule, if adopted, would not be a major Federal action
significantly affecting the quality of the human environment and,
therefore, an environmental impact statement is not required. The NRC's
generic determination in this regard is reflected in Sec. 51.32(b)(1).
The Commission has determined in Sec. 51.32 that there is no
significant environmental impact associated with the issuance of a
standard design certification or a design certification amendment, as
applicable.
The NRC's generic determination in this regard, as discussed in the
2007 final rule amending 10 CFR parts 51 and 52 (72 FR 49351; August
28, 2007), is based upon consideration that a design certification rule
does not authorize the siting, construction, or operation of a facility
referencing any particular design; it only codifies the NuScale design
in a rule. The NRC will evaluate the environmental impacts and issue an
environmental impact statement as appropriate under NEPA as part of the
application for the construction and operation of a facility
referencing any particular design certification rule.
Consistent with Sec. Sec. 51.30(d) and 51.32(b), the NRC has
prepared an environmental assessment for the NuScale design addressing
various design alternatives to prevent and mitigate severe accidents.
The environmental assessment is based, in part, upon the NRC's review
of NuScale Power's evaluation of various design alternatives to prevent
and mitigate severe accidents in Revision 5 of the DCA Part 3,
``Application Applicant's Environmental Report--Standard Design
Certification.'' Based on a review of NuScale Power's evaluation, the
NRC concludes that (1) NuScale Power identified a reasonably complete
set of potential design alternatives to prevent and mitigate severe
accidents for the NuScale design and (2) none of the potential design
alternatives appropriate at the design certification stage are
justified on the basis of cost-benefit considerations. These issues are
considered resolved for the NuScale design.
Based on its own independent evaluation, the NRC concluded that
none of the possible candidate design alternatives appropriate at this
design certification stage are potentially cost beneficial for NuScale
for accident events. This independent evaluation was based on
reasonable treatment of costs, benefits, and sensitivities. The NRC's
conclusion is applicable for sites with site characteristics that fall
within the site parameters of the representative site specified in the
NuScale environmental report. The NRC concludes that NuScale Power has
adequately identified areas appropriate at this design certification
stage where risk potentially could be reduced in a cost beneficial
manner and that NuScale Power has adequately assessed whether the
implementation of the identified potential severe accident mitigation
design alternatives (SAMDAs) or candidate design alternatives would be
cost beneficial for the representative site. As noted in the
environmental assessment, SAMDA candidates for multi-unit sites are
evaluated in the context of multiple NuScale reactor buildings, each
with up to 12 power modules at the same site. Site-specific SAMDAs,
multi-unit aspects, procedural and training SAMDAs, and the design
element details of the reactor building crane will need to be assessed
when an application for a specific site is submitted to construct and
operate a NuScale power plant.
The determination of this environmental assessment is that there
will be no significant offsite impact to the public from this action.
The environmental assessment is available as indicated under Section
XVIII of this document.
XIII. Paperwork Reduction Act
This final rule contains new or amended collections of information
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). The collections of information were approved by the Office of
Management and Budget, approval number 3150-0151.
The burden to the public for the information collections is
estimated to average 130 hours per response, including the time for
reviewing instructions, searching existing data sources, gathering and
maintaining the data needed, and completing and reviewing the
information collection.
The information collection is being conducted to fulfill the
requirements of a future applicant that references the design
certification to maintain records of changes to and departures from the
generic DCD, which are to be reflected in the plant-specific DCD. This
information will be used by the NRC to fulfill its responsibilities in
the licensing of nuclear power plants. Responses to this collection of
information are mandatory. Confidential and proprietary information
submitted to the NRC is protected in accordance with NRC regulations at
Sec. Sec. 9.17(a) and 2.39(b).
You may submit comments on any aspect of the information
collections, including suggestions for reducing the burden, by the
following methods:
Federal rulemaking website: Go to https://www.regulations.gov search for Docket ID NRC-2017-0029.
Mail comments to: FOIA, Library, and Information
Collections Branch, Office of the Chief Information Officer, Mail Stop:
T6-A10M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001
or to the OMB reviewer at: OMB Office of Information and Regulatory
Affairs (3150-0151), Attn: Desk Officer for the Nuclear Regulatory
Commission, 725 17th Street NW, Washington, DC 20503; email:
[email protected].
Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, a collection of information unless the document requesting
or requiring the collection displays a currently valid OMB control
number.
XIV. Congressional Review Act
This final rule is a rule as defined in the Congressional Review
Act (5 U.S.C. 801-808). However, the Office of Management and Budget
has not found it to be a major rule as defined in the Congressional
Review Act.
XV. Agreement State Compatibility
Under the ``Agreement State Program Policy Statement'' approved by
the Commission on October 2, 2017, and published in the Federal
Register on October 18, 2017 (82 FR 48535), this rule is classified as
compatibility ``NRC.'' Compatibility is not required for Category
``NRC'' regulations. The NRC program elements in this category are
those that relate directly to areas of regulation reserved to the NRC
by the AEA or the provisions of title 10 of the Code of Federal
Regulations, and although an Agreement State may not adopt program
elements reserved to the NRC, it may wish to inform its licensees of
certain requirements via a mechanism that is consistent with a
particular State's administrative procedure laws, but does not confer
regulatory authority on the State.
[[Page 3303]]
XVI. Voluntary Consensus Standards
The National Technology Transfer and Advancement Act of 1995,
Public Law 104-113, requires that Federal agencies use technical
standards that are developed or adopted by voluntary consensus
standards bodies unless the use of such a standard is inconsistent with
applicable law or otherwise impractical. In this final rule, the NRC
certifies the NuScale standard design for use in nuclear power plant
licensing under 10 CFR parts 50 or 52. Design certifications are not
generic rulemakings establishing a generally applicable standard with
which all 10 CFR parts 50 and 52 nuclear power plant licensees must
comply. Design certifications are Commission approvals of specific
nuclear power plant designs by rulemaking. Furthermore, design
certifications are initiated by an applicant for rulemaking, rather
than by the NRC. This action does not constitute the establishment of a
standard that contains generally applicable requirements.
XVII. Availability of Documents
The documents identified in the following table are available to
interested persons through one or more of the following methods, as
indicated.
Documents Related to NuScale Design Certification Rule
------------------------------------------------------------------------
ADAMS accession No./web
Document link/Federal Register
citation
------------------------------------------------------------------------
SECY-22-0062, ``Final Rule: NuScale Small ML22004A002
Modular Reactor Design Certification (RIN
3150-AJ98; NRC-2017-0029),'' July 1, 2022.
SECY-21-0004, ``Proposed Rule: NuScale Small ML19353A003
Modular Reactor Design Certification (RIN
3150-AJ98; NRC-2017-0029),'' January 14,
2021.
Staff Requirements Memorandum for SECY-21- ML21126A153
0004, ``Proposed Rule: NuScale Small Modular
Reactor Design Certification (RIN 3150-AJ98;
NRC-2017-0029),'' May 6, 2021.
Annotated Comment Submissions on Proposed ML22045A21
Rule: NuScale Small Modular Reactor Design
Certification (NRC-2017-0029; RIN 3150-
AJ98), June 2022.
Final Rule Comment Response Document for ML22216A015
NuScale Small Modular Reactor Design
Certification (public comment analysis
document), July 2022.
NuScale Power, LLC, Submittal of the NuScale ML20225A071
Standard Plant Design Certification
Application, Revision 5, July 2020.
NuScale Standard Design Certification ML20224A512
Application, Part 3, ``Applicant's
Environmental Report--Standard Design
Certification,'' Revision 5, July 2020.
NuScale Power, LLC, Submittal of the NuScale ML20205L562
Standard Plant Design Certification
Application, Revision 4.1, June 19, 2020.
NuScale Power, LLC, Submittal of the NuScale ML19241A431
Standard Plant Design Certification
Application, Part 2, Tier 2, Revision 3,
August 2019.
NuScale Power, LLC, Submittal of the NuScale ML18310A345
Standard Plant Design Certification
Application, Part 2, Tier 2, Revision 2,
October 2018.
NuScale Power, LLC, Topical report TR-0915- ML19112A172
17565, Revision 3, Accident Source Term
Methodology, April 21, 2019.
Proposed Rule for the NuScale Small Modular 86 FR 34999
Reactor Design Certification, July 1, 2021.
Extension of Comment Period for the Proposed 86 FR 47251
Rule, August 24, 2021.
Docketing Notice for the NuScale Power, LLC, 82 FR 15717
Design Certification Application (DCA),
March 30, 2017.
Notification of Receipt of the NuScale Power, 82 FR 11372
LLC, Design Certification Application (DCA),
February 22, 2017.
NuScale Power, LLC, Submittal of the NuScale ML17013A229
Standard Plant Design Certification
Application (NRC Project No. 0769), Revision
0, December 2016.
NuScale Power, LLC, Submittal of NuScale ML15258A846
Preliminary Concept of Operations Summary
and Response to NRC Questions on Control
Room Activities, September 15, 2015.
Information on Differing Professional Opinion ML22122A116
(DPO) 2020-004, May 13, 2022.
------------------------------------------------------------------------
Final Safety Evaluation Report and Supporting Documents
------------------------------------------------------------------------
NuScale DCA Final Safety Evaluation Report, ML20023A318
August 2020.
NRC Safety Evaluation for NuScale Power, LLC, ML20044E199
Topical Report, TR-0516-49422, ``Loss-of-
Coolant,'' Revision 1, November 2019.
NRC Safety Evaluation for NuScale Power, LLC, ML17340A524
Topical Report, TR-0815-16497, Revision 1,
``Safety Classification of Passive Nuclear
Power Plant Electrical Systems,'' December
13, 2017.
NRC Safety Evaluation for NuScale Power, LLC, ML19297G520
Topical Report, TR-0915-17565, Rev. 3,
``Accident Source Term Methodology,''
October 24, 2019.
NRR Response to Taskings in EDO DPO Appeal ML22062A007
Decision Concerning DPO-2020-004, May 13,
2022.
------------------------------------------------------------------------
Environmental Reviews
------------------------------------------------------------------------
Final Environmental Assessment by the U.S. ML22216A014
Nuclear Regulatory Commission Relating to
the Certification of the NuScale Standard
Design, July 2022.
Environmental Assessment by the U.S. Nuclear ML19303C179
Regulatory Commission Relating to the
Certification of the NuScale Standard
Design, January 14, 2021.
Staff Technical Analysis in Support of the ML19302E819
NuScale Design Certification Environmental
Assessment, August 4, 2020.
------------------------------------------------------------------------
[[Page 3304]]
Commission Papers, Staff Requirement Memoranda, and Other Supporting
Documents
------------------------------------------------------------------------
SECY-11-0098, ``Operator Staffing for Small ML111870574
or Multi-Module Nuclear Power Plant
Facilities,'' July 22, 2011.
SECY-17-0075, ``Planned Improvements in ML16196A321
Design Certification Tiered Information
Designations,'' dated July 24, 2017.
SECY-18-0099, ``NuScale Power Exemption ML18065A431
Request from 10 CFR Part 50, Appendix A,
General Design Criterion 27, `Combined
Reactivity Control Systems Capability,' ''
dated October 9, 2018.
SECY-19-0079, ``Staff Approach to Evaluate ML19107A455
Accident Source Terms for the NuScale Power
Design Certification Application,'' August
16, 2019.
SECY-77-439, ``Single Failure Criterion,'' ML060260236
August 17, 1977.
SECY-93-087, ``Policy, Technical, and ML003708021
Licensing Issues Pertaining to Evolutionary
and Advanced Light-Water Reactor (ALWR)
Designs,'' April 2, 1993.
SRM-SECY-19-0036, ``Staff Requirements--SECY- ML19183A408
19-0036--Application of the Single Failure
Criterion to NuScale Power LLC's Inadvertent
Actuation Block Valves,'' July 2, 2019.
SRM-SECY-94-084, ``Policy and Technical ML003708098
Issues associated with the Regulatory
Treatment of Non-Safety Systems and
Implementation of Design Certification and
Light-Water Reactor Design Issues,'' June
30, 1994.
SRM-SECY-90-377, ``Requirements for Design ML003707892
Certification under 10 CFR part 52,''
February 15, 1991.
Response to NuScale Power, LLC Key Issue ML16229A522
Resolution Letter, Supplemental Response
Regarding Multi-Module Questions, October
25, 2016.
Advisory Committee on Reactor Safeguards ML20211M386
(ACRS) Letter, ``Report on the Safety
Aspects of the NuScale Small Modular
Reactor,'' July 29, 2020.
American Society of Mechanical Engineers https://webstore.ansi.org/
Standard QME-1-2007, ``Qualification of standards/asme/
Active Mechanical Equipment Used in Nuclear ansiasmeqme2007
Power Plants,'' 2007.
NRC Regulatory Guide 1.100, Rev. 3, ``Seismic ML091320468
Qualification of Electrical and Active
Mechanical Equipment and Functional
Qualification of Active Mechanical Equipment
for Nuclear Power Plants,'' September 2009.
NRC Regulatory Guide 1.206, Rev. 1, ML18131A181
``Applications for Nuclear Power Plants,''
October 2018.
NRC Agreement State Program Policy Statement, 82 FR 48535
October 18, 2017.
Final Rule for Licenses, Certifications, and 72 FR 49351
Approvals for Nuclear Power Plants (10 CFR
parts 51 and 52), August 28, 2007.
Office of the Federal Register (OFR) Final 79 FR 66267
Rule for Incorporation by Reference,
November 7, 2014.
Presidential Memorandum, ``Plain Language in 63 FR 31883
Government Writing,'' June 10, 1998.
Regulatory History of Design Certification, ML003761550
April 2000 \2\.
------------------------------------------------------------------------
NuScale Technical and Topical Reports
------------------------------------------------------------------------
ES-0304-1381-NP, Human-System Interface Style ML19338E948
Guide, Rev. 4, December 2019.
RP-0215-10815-NP, Concept of Operations, Rev. ML19133A293
3, May 2019.
RP-0316-17614-NP, Human Factors Engineering ML16364A342
Operating Experience Review Results Summary
Report, Rev. 0, December 2016 \3\.
RP-0316-17615-NP, Human Factors Engineering ML16364A342
Functional Requirements Analysis and
Function Allocation Results Summary Report,
Rev. 0, December 2016 \3\.
RP-0316-17616-NP, Human Factors Engineering ML19119A393
Task Analysis Results Summary Report, Rev.
2, April 2019.
RP-0316-17617-NP, Human Factors Engineering ML17004A222
Staffing and Qualifications Results Summary
Report, Rev. 0, December 2016 \3\.
RP-0316-17618-NP, Human Factors Engineering ML17004A222
Treatment of Important Human Actions Results
Summary Report, Rev. 0, December 2016 \3\.
RP-0316-17619-NP, Human Factors Engineering ML19119A398
Human-System Interface Design Results
Summary Report, Rev. 2, April 2019.
RP-0516-49116-NP, Control Room Staffing Plan ML16364A356
Validation Results, Rev. 1, December 2016.
RP-0914-8534-NP, Human Factors Engineering ML19119A342
Program Management Plan, Rev. 5, April 2019.
RP-0914-8543-NP, Human Factors Verification ML19119A372
and Validation Implementation Plan, Rev. 5,
April 2019.
RP-0914-8544-NP, Human Factors Engineering ML19331A910
Design Implementation Plan, Rev. 4, November
2019.
RP-1018-61289-NP, Human Factors Engineering ML19212A773
Verification and Validation Results Summary
Report, Rev. 1, July 2019.
RP-1215-20253-NP, Control Room Staffing Plan ML16364A353
Validation Methodology, Rev. 3, December
2016.
TR-0116-20781-NP, Fluence Calculation ML19183A485
Methodology and Results, Rev. 1, July 2019.
TR-0116-20825-NP-A, Applicability of AREVA ML18040B306
Fuel Methodology for the NuScale Design,
Rev. 1, June 2016.
TR-0116-21012-NP-A, NuScale Power Critical ML18360A632
Heat Flux Correlations, Rev. 1, December
2018.
TR-0316-22048-NP, Nuclear Steam Supply System ML20141M764
Advanced Sensor Technical Report, Rev. 3,
May 2020.
TR-0515-13952-NP-A, Risk Significance ML16284A016
Determination, Rev. 0, October 2016.
TR-0516-49084-NP, Containment Response ML20141L808
Analysis Methodology Technical Report, Rev.
3, May 2020.
[[Page 3305]]
TR-0516-49416-NP-A, Non-Loss-of-Coolant ML20191A281
Accident Analysis Methodology, Rev. 3, July
2020.
TR-0516-49417-NP-A, Evaluation Methodology ML20078Q094
for Stability Analysis of the NuScale Power
Module, Rev. 1, March 2020.
TR-0516-49422-NP-A, Loss-of-Coolant Accident ML20189A644
Evaluation Model, Rev. 2, July 2020.
TR-0616-48793-NP-A, Nuclear Analysis Codes ML18348B036
and Methods Qualification, Rev. 1, November
2018.
TR-0616-49121-NP, NuScale Instrument Setpoint ML20141M114
Methodology Technical Report, Rev. 3, May
2020.
TR-0716-50350-NP-A, Rod Ejection Accident ML20168B203
Methodology, Rev. 1, June 2020.
TR-0716-50351-NP-A, NuScale Applicability of ML20122A248
AREVA Method for the Evaluation of Fuel
Assembly Structural Response to Externally
Applied Forces, Rev. 1, April 2020.
TR-0716-50424-NP, Combustible Gas Control, ML19091A232
Rev. 1, March 2019.
TR-0716-50439-NP, NuScale Comprehensive ML19212A776
Vibration Assessment Program Analysis
Technical Report, Rev. 2, July 2019.
TR-0815-16497-NP-A, Safety Classification of ML18054B607
Passive Nuclear Power Plant Electrical
Systems Topical Report, Rev. 1, January 2018.
TR-0816-49833-NP, Fuel Storage Rack Analysis, ML18310A154
Rev. 1, November 2018.
TR-0816-50796-NP, Loss of Large Areas Due to ML19165A294
Explosions and Fires Assessment, Rev. 1,
June 2019.
TR-0816-50797 (NuScale Nonproprietary), ML19302H598
Mitigation Strategies for Loss of All AC
Power Event, Rev. 3, October 2019.
TR-0816-51127-NP, NuFuel-HTP2TM Fuel and ML19353A719
Control Rod Assembly Designs, Rev. 3,
December 2019.
TR-0818-61384-NP, Pipe Rupture Hazards ML19212A682
Analysis, Rev. 2, July 2019.
TR-0915-17564-NP-A, Subchannel Analysis ML19067A256
Methodology, Rev. 2, February 2019.
TR-0915-17565-NP-A, Accident Source Term ML20057G132
Methodology, Rev. 4, February 2020.
TR-0916-51299-NP, Long-Term Cooling ML20141L816
Methodology, Rev. 3, May 2020.
TR-0916-51502-NP, NuScale Power Module ML19093B850
Seismic Analysis, Rev. 2, April 2019.
TR-0917-56119-NP, CNV Ultimate Pressure ML19158A382
Integrity, Rev. 1, June 2019.
TR-0918-60894-NP, Comprehensive Vibration ML19214A248
Assessment Program Measurement and
Inspection Plan Technical Report, Rev. 1,
August 2019.
TR-1010-859-NP-A, NuScale Topical Report: ML20176A494
Quality Assurance Program Description for
the NuScale Power Plant, Rev. 5, May 2020.
TR-1015-18177-NP, Pressure and Temperature ML18298A304
Limits Methodology, Rev. 2, October 2018.
TR-1015-18653-NP-A, Design of the Highly ML17256A892
Integrated Protection System Platform
Topical Report, Rev. 2, May 2017.
TR-1016-51669-NP, NuScale Power Module Short- ML19211D411
Term Transient Analysis, Rev. 1, July 2019.
TR-1116-51962-NP, NuScale Containment Leakage ML19149A298
Integrity Assurance, Rev. 1, May 2019.
TR-1116-52065-NP, Effluent Release (GALE ML18317A364
Replacement) Methodology and Results, Rev.
1, November 2018.
------------------------------------------------------------------------
The NRC may post materials related to this document, including
public comments, on the Federal rulemaking website at https://www.regulations.gov under Docket ID NRC-2017-0029. In addition, the
Federal rulemaking website allows members of the public to receive
alerts when changes or additions occur in a docket folder. To
subscribe: (1) navigate to the docket folder (NRC-2017-0029); (2) click
the ``Subscribe'' link; and (3) enter an email address and click on the
``Subscribe'' link.
---------------------------------------------------------------------------
\2\ The regulatory history of the NRC's design certification
reviews is a package of documents that is available in the NRC's PDR
and NRC Library. This history spans the period during which the NRC
simultaneously developed the regulatory standards for reviewing
these designs and the form and content of the rules that certified
the designs.
\3\ The duplicate ADAMS Accession Nos. ML16364A342 and
ML17004A222 are intentional and indicate when multiple reports are
part of a single submittal.
---------------------------------------------------------------------------
XVIII. Incorporation by Reference--Reasonable Availability to
Interested Parties
The NRC is incorporating by reference the NuScale DCA, Revision 5.
As described in the ``Discussion'' sections of this document, the
generic DCD includes Tier 1 and Tier 2 information (including the
technical and topical reports referenced in Chapter 1) and generic
technical specifications in order to effectively control this
information and facilitate its incorporation by reference into the
rule. NuScale Power submitted Revision 5 of the DCA to the NRC in July
2020.
The NRC is required by law to obtain approval for incorporation by
reference from the Office of the Federal Register (OFR). The OFR's
requirements for incorporation by reference are set forth in 1 CFR part
51. On November 7, 2014, the OFR adopted changes to its regulations
governing incorporation by reference (79 FR 66267). The OFR regulations
require an agency to discuss, in the preamble of the final rule, the
ways that the materials it incorporates by reference are reasonably
available to interested parties and how interested parties can obtain
the materials. The discussion in this section complies with the
requirement for final rules as set forth in 1 CFR 51.5(a)(1).
The NRC considers ``interested parties'' to include all potential
NRC stakeholders, not only the individuals and entities regulated or
otherwise subject to the NRC's regulatory oversight. These NRC
stakeholders are not a homogenous group but vary with respect to the
considerations for determining reasonable availability. Therefore, the
NRC distinguishes between different classes of interested parties for
the purposes of determining whether the material is ``reasonably
available.'' The NRC considers the following to be classes of
interested parties in NRC rulemakings with regard
[[Page 3306]]
to the material to be incorporated by reference:
Individuals and small entities regulated or otherwise
subject to the NRC's regulatory oversight (this class also includes
applicants and potential applicants or licenses and other NRC
regulatory approvals) and who are subject to the material to be
incorporated by reference by rulemaking. In this context, ``small
entities'' has the same meaning as a ``small entity'' under Sec.
2.810.
Large entities otherwise subject to the NRC's regulatory
oversight (this class also includes applicants and potential applicants
for licenses and other NRC regulatory approvals) and who are subject to
the material to be incorporated by reference by rulemaking. In this
context, ``large entities'' are those which do not qualify as a ``small
entity'' under Sec. 2.810.
Non-governmental organizations with institutional
interests in the matters regulated by the NRC.
Other Federal agencies, States, and local governmental
bodies (within the meaning of Sec. 2.315(c)).
Federally-recognized and State-recognized \4\ Indian
tribes.
---------------------------------------------------------------------------
\4\ State-recognized Indian tribes are not within the scope of
10 CFR 2.315(c). However, for purposes of the NRC's compliance with
1 CFR 51.5, ``interested parties'' includes a broad set of
stakeholders, including State-recognized Indian tribes.
---------------------------------------------------------------------------
Members of the general public (i.e., individual,
unaffiliated members of the public who are not regulated or otherwise
subject to the NRC's regulatory oversight) who may wish to gain access
to the materials which the NRC incorporates by reference by rulemaking
in order to participate in the rulemaking process.
The NRC makes the materials incorporated by reference available for
inspection to all interested parties, by appointment, at the NRC
Technical Library, which is located at Two White Flint North, 11545
Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-7000;
email: [email protected]. In addition, as described in Section
XVIII of this document, documents related to this final rule are
available online in the NRC's ADAMS Public Documents collection at
https://www.nrc.gov/reading-rm/adams.html.
The NRC concludes that the materials the NRC is incorporating by
reference in this final rule are reasonably available to all interested
parties because the materials are available in multiple ways and in a
manner consistent with their interest in the materials.
List of Subjects in 10 CFR Part 52
Administrative practice and procedure, Antitrust, Combined license,
Early site permit, Emergency planning, Fees, Incorporation by
reference, Inspection, Issue finality, Limited work authorization,
Nuclear power plants and reactors, Probabilistic risk assessment,
Prototype, Reactor siting criteria, Redress of site, Penalties,
Reporting and recordkeeping requirements, Standard design, Standard
design certification.
For the reasons set out in the preamble and under the authority of
the Atomic Energy Act of 1954, as amended; the Energy Reorganization
Act of 1974, as amended; the Nuclear Waste Policy Act of 1982, as
amended; and 5 U.S.C. 552 and 553, the NRC is amending 10 CFR part 52
as follows:
PART 52--LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER
PLANTS
0
1. The authority citation for part 52 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 103, 104, 147, 149,
161, 181, 182, 183, 185, 186, 189, 223, 234 (42 U.S.C. 2133, 2134,
2167, 2169, 2201, 2231, 2232, 2233, 2235, 2236, 2239, 2273, 2282);
Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42
U.S.C. 5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.
Sec. 52.11 [Amended]
0
2. In Sec. 52.11(b), remove the phrase ``appendices A, B, C, D, E, F,
and N of this part'' and add, in its place, the phrase ``appendices A,
B, C, D, E, F, G, and N of this part''.
0
3. Add appendix G to part 52 to read as follows:
Appendix G to Part 52--Design Certification Rule for NuScale
I. Introduction
Appendix G constitutes the standard design certification for the
NuScale design (hereinafter referred to as NuScale), in accordance
with 10 CFR part 52, subpart B. The applicant for this standard
design certification NuScale is NuScale Power, LLC.
II. Definitions
A. Generic design control document (generic DCD) means the
documents containing the Tier 1 and Tier 2 information (including
the technical and topical reports referenced in Chapter 1) and
generic technical specifications that are incorporated by reference
into this appendix.
B. Generic technical specifications (generic TS) means the
information required by 10 CFR 50.36 and 50.36a for the portion of
the plant that is within the scope of this appendix.
C. Plant-specific DCD means that portion of the combined license
(COL) final safety analysis report (FSAR) that sets forth both the
generic DCD information and any plant-specific changes to generic
DCD information.
D. Tier 1 means the portion of the design-related information
contained in the generic DCD that is approved and certified by this
appendix (Tier 1 information). The design descriptions, interface
requirements, and site parameters are derived from Tier 2
information. Tier 1 information includes:
1. Definitions and general provisions;
2. Design descriptions;
3. Inspections, tests, analyses, and acceptance criteria
(ITAAC);
4. Significant site parameters; and
5. Significant interface requirements.
E. Tier 2 means the portion of the design-related information
contained in the generic DCD that is approved but not certified by
this appendix (Tier 2 information). Compliance with Tier 2 is
required, but generic changes to and plant-specific departures from
Tier 2 are governed by Section VIII of this appendix. Compliance
with Tier 2 provides a sufficient, but not the only acceptable,
method for complying with Tier 1. Compliance methods differing from
Tier 2 must satisfy the change process in Section VIII of this
appendix. Regardless of these differences, an applicant or licensee
must meet the requirement in paragraph III.B of this appendix to
reference Tier 2 when referencing Tier 1. Tier 2 information
includes:
1. Information required by Sec. 52.47(a) and (c), with the
exception of generic TS and conceptual design information;
2. Supporting information on the inspections, tests, and
analyses that will be performed to demonstrate that the acceptance
criteria in the ITAAC have been met; and
3. COL action items (COL license information) identify certain
matters that must be addressed in the site-specific portion of the
FSAR by an applicant who references this appendix. These items
constitute information requirements but are not the only acceptable
set of information in the FSAR. An applicant may depart from or omit
these items, provided that the departure or omission is identified
and justified in the FSAR. After issuance of a construction permit
or COL, these items are not requirements for the licensee unless
such items are restated in the FSAR.
F. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety
analyses means:
1. Changing any of the elements of the method described in the
plant-specific DCD unless the results of the analysis are
conservative or essentially the same; or
2. Changing from a method described in the plant-specific DCD to
another method unless that method has been approved by the NRC for
the intended application.
G. Nuclear power unit, as applied to this certified design,
means a nuclear power module and associated equipment necessary for
electric power generation and includes those structures, systems,
and components required to provide reasonable assurance the facility
can be operated without undue risk to the health and safety of the
public.
H. All other terms in this appendix have the meaning set out in
10 CFR 50.2, 10 CFR
[[Page 3307]]
52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as
applicable.
III. Scope and Contents
A. Incorporation by reference.
1. Certain material listed in paragraph III.A.2 of this appendix
is incorporated by reference into this appendix G with the approval
of the Director of the Federal Register in accordance with 5 U.S.C.
552(a) and 1 CFR part 51. All approved incorporation by reference
(IBR) material in paragraph III.A.2 of this appendix may be obtained
from NuScale Power, LLC, 6650 SW Redwood Lane, Suite 210, Portland,
Oregon 97224, telephone: 1-971-371-1592, email:
[email protected], and can be inspected as follows:
a. Contact the U.S. Nuclear Regulatory Commission at: U.S.
Nuclear Regulatory Commission, Two White Flint North, 11545
Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-7000;
email: [email protected]; https://www.nrc.gov/reading-rm/pdr.html.
b. Access ADAMS and view the material online in the NRC Library
at https://www.nrc.gov/reading-rm/adams.html. In ADAMS, search under
ADAMS Accession No. ML20225A071. The material is available in the
ADAMS Public Documents collection.
c. If you do not have access to ADAMS or if you have problems
accessing documents located in ADAMS, contact the NRC's Public
Document Room (PDR) reference staff at 1-800-397-4209, 301-415-3747,
or by email at [email protected].
d. For information on the availability of this material at the
National Archives and Records Administration, visit
www.archives.gov/federal-register/cfr/ibr-locations.html or email:
[email protected].
2. Material incorporated by reference.
a. NuScale Standard Plant Design Certification Application,
Certified Design Descriptions and Inspections, Tests, Analyses, &
Acceptance Criteria (ITAAC), Part 2--Tier 1, Revision 5, July 2020.
b. NuScale Standard Plant Design Certification Application, Part
2--Tier 2, Revision 5, July 2020, including:
i. Chapter One, Introduction and General Description of the
Plant.
ii. Chapter Two, Site Characteristics and Site Parameters.
iii. Chapter Three, Design of Structures, Systems, Components
and Equipment.
iv. Chapter Four, Reactor.
v. Chapter Five, Reactor Coolant System and Connecting Systems.
vi. Chapter Six, Engineered Safety Features.
vii. Chapter Seven, Instrumentation and Controls.
viii. Chapter Eight, Electric Power.
ix. Chapter Nine, Auxiliary Systems.
x. Chapter Ten, Steam and Power Conversion System.
xi. Chapter Eleven, Radioactive Waste Management.
xii. Chapter Twelve, Radiation Protection.
xiii. Chapter Thirteen, Conduct of Operations.
xiv. Chapter Fourteen, Initial Test Program and Inspections,
Tests, Analyses, and Acceptance Criteria.
xv. Chapter Fifteen, Transient and Accident Analyses.
xvi. Chapter Sixteen, Technical Specifications.
xvii. Chapter Seventeen, Quality Assurance and Reliability
Assurance.
xviii. Chapter Eighteen, Human Factors Engineering.
xix. Chapter Nineteen, Probabilistic Risk Assessment and Severe
Accident Evaluation.
xx. Chapter Twenty, Mitigation of Beyond-Design-Basis Events.
xxi. Chapter Twenty-One, Multi-Module Design Considerations.
c. DCA Part 4, Volume 1, Revision 5.0, Generic Technical
Specifications, NuScale Nuclear Power Plants, Volume 1:
Specifications.
d. DCA Part 4, Volume 2, Revision 5.0, Generic Technical
Specifications, NuScale Nuclear Power Plants, Volume 2: Bases.
e. ES-0304-1381-NP, Human-System Interface Style Guide, December
2019, Revision 4.
f. RP-0215-10815-NP, Concept of Operations, May 2019, Revision
3.
g. RP-0316-17614-NP, Human Factors Engineering Operating
Experience Review Results Summary Report, December 7, 2016, Revision
0.
h. RP-0316-17615-NP, Human Factors Engineering Functional
Requirements Analysis and Function Allocation Results Summary
Report, December 2, 2016, Revision 0.
i. RP-0316-17616-NP, Human Factors Engineering Task Analysis
Results Summary Report, April 2019, Revision 2.
j. RP-0316-17617-NP, Human Factors Engineering Staffing and
Qualifications Results Summary Report, December 2, 2016, Revision 0.
k. RP-0316-17618-NP, Human Factors Engineering Treatment of
Important Human Actions Results Summary Report, December 2, 2016,
Revision 0.
l. RP-0316-17619-NP, Human Factors Engineering Human-System
Interface Design Results Summary Report, April 2019, Revision 2.
m. RP-0516-49116-NP, Control Room Staffing Plan Validation
Results, December 2, 2016, Revision 1.
n. RP-0914-8534-NP, Human Factors Engineering Program Management
Plan, April 2019, Revision 5.
o. RP-0914-8543-NP, Human Factors Verification and Validation
Implementation Plan, April 2019, Revision 5.
p. RP-0914-8544-NP, Human Factors Engineering Design
Implementation Plan, November 2019, Revision 4.
q. RP-1018-61289-NP, Human Factors Engineering Verification and
Validation Results Summary Report, July 2019, Revision 1.
r. RP-1215-20253-NP, Control Room Staffing Plan Validation
Methodology, December 2, 2016, Revision 3.
s. TR-0116-20781-NP, Fluence Calculation Methodology and
Results, July 2019, Revision 1.
t. TR-0116-20825-NP-A, Applicability of AREVA Fuel Methodology
for the NuScale Design, June 2016, Revision 1.
u. TR-0116-21012-NP-A, NuScale Power Critical Heat Flux
Correlations, December 2018, Revision 1.
v. TR-0316-22048-NP, Nuclear Steam Supply System Advanced Sensor
Technical Report, May 2020, Revision 3.
w. TR-0515-13952-NP-A, Risk Significance Determination, October
2016, Revision 0.
x. TR-0516-49084-NP, Containment Response Analysis Methodology
Technical Report, May 2020, Revision 3.
y. TR-0516-49416-NP-A, Non-Loss-of-Coolant Accident Analysis
Methodology, July 2020, Revision 3.
z. TR-0516-49417-NP-A, Evaluation Methodology for Stability
Analysis of the NuScale Power Module, March 2020, Revision 1.
aa. TR-0516-49422-NP-A, Loss-of-Coolant Accident Evaluation
Model, July 2020, Revision 2.
ab. TR-0616-48793-NP-A, Nuclear Analysis Codes and Methods
Qualification, November 2018, Revision 1.
ac. TR-0616-49121-NP, NuScale Instrument Setpoint Methodology
Technical Report, May 2020, Revision 3.
ad. TR-0716-50350-NP-A, Rod Ejection Accident Methodology, June
2020, Revision 1.
ae. TR-0716-50351-NP-A, NuScale Applicability of AREVA Method
for the Evaluation of Fuel Assembly Structural Response to
Externally Applied Forces, April 2020, Revision 1.
af. TR-0716-50424-NP, Combustible Gas Control, March 2019,
Revision 1.
ag. TR-0716-50439-NP, NuScale Comprehensive Vibration Assessment
Program Analysis Technical Report, July 2019, Revision 2.
ah. TR-0815-16497-NP-A, Safety Classification of Passive Nuclear
Power Plant Electrical Systems, January 2018, Revision 1.
ai. TR-0816-49833-NP, Fuel Storage Rack Analysis, November 2018,
Revision 1.
aj. TR-0816-50796-NP, Loss of Large Areas Due to Explosions and
Fires Assessment, June 2019, Revision 1.
ak. TR-0816-50797, Mitigation Strategies for Loss of All AC
Power Event [NuScale Nonproprietary], October 2019, Revision 3.
al. TR-0816-51127-NP, NuFuel-HTP2TM Fuel and Control
Rod Assembly Designs, December 2019, Revision 3.
am. TR-0818-61384-NP, Pipe Rupture Hazards Analysis, July 2019,
Revision 2.
an. TR-0915-17564-NP-A, Subchannel Analysis Methodology,
February 2019, Revision 2.
ao. TR-0915-17565-NP-A, Accident Source Term Methodology,
February 2020, Revision 4.
ap. TR-0916-51299-NP, Long-Term Cooling Methodology, May 2020,
Revision 3.
aq. TR-0916-51502-NP, NuScale Power Module Seismic Analysis,
April 2019, Revision 2.
ar. TR-0917-56119-NP, CNV Ultimate Pressure Integrity, June
2019, Revision 1.
as. TR-0918-60894-NP, NuScale Comprehensive Vibration Assessment
Program Measurement and Inspection Plan Technical Report, August
2019, Revision 1.
at. NP-TR-1010-859-NP-A, NuScale Topical Report: Quality
Assurance Program
[[Page 3308]]
Description for the NuScale Power Plant, May 2020, Revision 5.
au. TR-1015-18177-NP, Pressure and Temperature Limits
Methodology, October 2018, Revision 2.
av. TR-1015-18653-NP-A, Design of the Highly Integrated
Protection System Platform, May 2017, Revision 2.
aw. TR-1016-51669-NP, NuScale Power Module Short-Term Transient
Analysis, July 2019, Revision 1.
ax. TR-1116-51962-NP, NuScale Containment Leakage Integrity
Assurance, May 2019, Revision 1.
ay. TR-1116-52065-NP, Effluent Release (GALE Replacement)
Methodology and Results, November 2018, Revision 1.
B.1. An applicant or licensee referencing this appendix, in
accordance with Section IV of this appendix, shall incorporate by
reference and comply with the requirements of this appendix except
as otherwise provided in this appendix.
2. Conceptual design information, as set forth in the design
certification application Part 2, Tier 2, Section 1.2, and the
discussion of ``first principles'' contained in design certification
application Part 2, Tier 2, Section 14.3.2, are not incorporated by
reference into this appendix.
C. If there is a conflict between Tier 1 and Tier 2 of the DCD,
then Tier 1 controls.
D. If there is a conflict between the generic DCD and either the
application for the design certification of NuScale or the final
safety evaluation report related to certification of the NuScale
standard design, then the generic DCD controls.
E. Design activities for structures, systems, and components
that are wholly outside the scope of this appendix may be performed
using site characteristics, provided the design activities do not
affect the DCD or conflict with the interface requirements.
IV. Additional Requirements and Restrictions
A. An applicant for a COL that wishes to reference this appendix
shall, in addition to complying with the requirements of Sec. Sec.
52.77, 52.79, and 52.80, comply with the following requirements:
1. Incorporate by reference, as part of its application, this
appendix.
2. Include, as part of its application:
a. A plant-specific DCD containing the same type of information
and using the same organization and numbering as the generic DCD for
NuScale, either by including or incorporating by reference the
generic DCD information, and as modified and supplemented by the
applicant's exemptions and departures;
b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
c. Plant-specific TS, consisting of the generic and site-
specific TS that are required by 10 CFR 50.36 and 50.36a;
d. Information demonstrating that the site characteristics fall
within the site parameters and that the interface requirements have
been met;
e. Information that addresses the COL action items;
f. Information required by Sec. 52.47(a) that is not within the
scope of this appendix;
g. Information demonstrating that necessary shielding to limit
radiological dose consistent with the radiation zones specified in
design certification application Part 2, Tier 2, Chapter 12, Figure
12.3-1, ``Reactor Building Radiation Zone Map,'' is provided to
account for penetrations in the radiation shield wall between the
power module bay and the reactor building steam gallery area;
h. Information demonstrating that the requirements of 10 CFR
50.34(f)(2)(xxviii) are met with respect to potential radiological
releases under accident conditions from the systems used for post-
accident hydrogen and oxygen monitoring described in design
certification application Part 2, Tier 2, Section 6.2.5; information
demonstrating that post-accident leakage from these systems does not
result in the total main control room dose exceeding the dose
criteria for the surrogate event with significant core damage, which
may include use of design features compliant with 10 CFR
50.34(f)(2)(vii), as appropriate; and information demonstrating that
post-accident leakage from these systems does not result in the
total dose for the surrogate event with significant core damage
exceeding the offsite dose criteria, as required by 10 CFR
52.47(a)(2)(iv); and
i. Information demonstrating that the requirements of 10 CFR
52.47(a)(2)(iv) and General Design Criterion (GDC) 4 and GDC 31 of
appendix A to 10 CFR part 50 are met with respect to the structural
and leakage integrity of the steam generator tubes that might be
compromised by effects from density wave oscillations in the
secondary fluid system, including the method of analysis to predict
the thermal-hydraulic conditions of the steam generator secondary
fluid system and resulting loads, stresses, and deformations from
density wave oscillations and reverse flow. This information must be
consistent with the other design information regarding steam
generator integrity contained in design certification application
Part 2, Tier 2, Sections 3.9.2 and 5.4.1.
3. Include, in the plant-specific DCD, the sensitive,
unclassified, non-safeguards information (including proprietary
information and security-related information) and safeguards
information referenced in the NuScale generic DCD.
4. Include, as part of its application, a demonstration that an
entity other than NuScale Power, LLC, is qualified to supply the
NuScale generic DCD, unless NuScale Power, LLC, supplies the design
for the applicant's use.
B. The Commission reserves the right to determine in what manner
this appendix may be referenced by an applicant for a construction
permit or operating license under 10 CFR part 50.
C. A licensee referencing the NuScale design certification is
exempt from portions of the following regulation:
1. Paragraph (m) of 10 CFR 50.54--Minimum Staffing. In lieu of
these requirements, a licensee that references this appendix must
comply with the following:
a. A senior operator licensed pursuant to part 55 of this
chapter shall be present at the facility or readily available on
call at all times during its operation, and shall be present at the
facility during initial startup and approach to power, recovery from
an unplanned or unscheduled shutdown or significant reduction in
power, and refueling, or as otherwise prescribed in the facility
license.
b. Licensees shall meet the following requirements:
i. Each licensee shall meet the minimum licensed operator
staffing requirements identified in Table 1:
Table 1--Minimum Requirements per Shift for On-Site Staffing of NuScale
Power Plants by Operators and Senior Operators Licensed Under 10 CFR
Part 55
------------------------------------------------------------------------
Number of units operating (a One to twelve
nuclear power unit is units
considered to be operating ---------------
when it is in MODE 1, 2, or Position
3 as defined by the unit's One control
technical specifications) room
------------------------------------------------------------------------
None........................ Senior operator........... 1
Operator.................. 2
One to twelve............... Senior operator........... 3
Operator.................. 3
------------------------------------------------------------------------
Source: Design Certification Application, Part 7, Section 6.1.3,
``Requested Action.''
ii. Each facility licensee shall have at its site a person
holding a senior operator license for all fueled units at the site
who is assigned responsibility for overall plant operation at all
times there is fuel in any unit. At all times any module is fueled,
regardless of mode, there must be a licensed operator or senior
operator in the control room.
iii. When a nuclear power unit is in MODE 1, 2, or 3, as defined
by the unit's technical
[[Page 3309]]
specifications, each licensee shall have a person holding a senior
operator license for the nuclear power unit in the control room at
all times. In addition to this senior operator, a second person who
is either a licensed operator or licensed senior operator shall be
present at the controls at all times. A third person who is either a
licensed operator or licensed senior operator shall be in the
control room envelope at all times.
iv. Each licensee shall have present, during alteration or
movement of the core of a nuclear power unit (including fuel
loading, fuel transfer, or movement of a module that contains fuel),
a person holding a senior operator license or a senior operator
license limited to fuel handling to directly supervise the activity
and, during this time, the licensee shall not assign other duties to
this person.
2. Appendix J to 10 CFR part 50, Type A testing--Primary Reactor
Containment Leakage Testing for Water-Cooled Power Reactors.
V. Applicable Regulations
A. Except as indicated in paragraph B of this section, the
regulations that apply to NuScale are in 10 CFR parts 20, 50, 52,
73, and 100, codified as of February 21, 2023, that are applicable
and technically relevant, as described in the final safety
evaluation report.
B. The NuScale design is exempt from portions of the following
regulations:
1. Paragraph (f)(2)(vi) of 10 CFR 50.34 and 10 CFR 50.46a--High
point venting for the reactor coolant system and reactor pressure
vessel head.
2. Paragraph (f)(2)(viii) of 10 CFR 50.34--Post-accident
sampling of the reactor coolant system and containment.
3. Paragraph (f)(2)(xiii) of 10 CFR 50.34--Power supplies for
pressurizer heaters.
4. Paragraph (f)(2)(xiv)(E) of 10 CFR 50.34--Automatic closing
of containment isolation systems on a high radiation signal.
5. Paragraph (f)(2)(xx) of 10 CFR 50.34--Power from vital buses
and emergency power sources for pressurizer level indication.
6. Paragraph (c)(2) of 10 CFR 50.44--Combustible gas control.
7. Paragraph (a)(1)(i) of 10 CFR 50.46--Applicability limited to
reactor designs that use zircaloy or ZIRLO fuel rod cladding
material.
8. Paragraph (c)(1) of 10 CFR 50.62--Diverse equipment to
initiate a turbine trip under conditions indicative of an
anticipated transient without scram.
9 Appendix A of 10 CFR part 50--Electric Power Systems GDCs:
a. GDC 17--Electric power systems for safety-related functions;
b. GDC 18--Design to permit periodic inspection and testing of
electric power systems;
c. GDC 34--Electric power systems for residual heat removal;
d. GDC 35--Electric power systems for emergency core cooling;
e. GDC 38--Electric power systems for containment heat removal;
f. GDC 41--Electric power systems for containment atmosphere
cleanup; and
g. GDC 44--Electric power systems for cooling.
10. Appendix A to 10 CFR part 50, GDC 19--Equipment outside the
control room with capability for cold shutdown of the reactor.
11. Appendix A to 10 CFR part 50, GDC 27--Demonstration of long-
term shutdown under post-accident conditions with an assumed worst
rod stuck out.
12. Appendix A to 10 CFR part 50, GDC 33--Reactor coolant makeup
for protection against small breaks in the reactor coolant pressure
boundary.
13. Appendix A to 10 CFR part 50, GDC 40--Periodic pressure and
functional testing of containment heat removal system.
14. Appendix A to 10 CFR part 50, GDC 52--Design to allow
periodic containment leakage rate testing.
15. Appendix A of 10 CFR part 50, GDCs 55, 56, and 57--
Containment Isolation:
a. GDC 55--Isolation valves for certain reactor coolant pressure
boundary lines penetrating containment;
b. GDC 56--Isolation valves for certain primary containment
lines; and
c. GDC 57--Isolation valves for certain closed systems lines.
16. Appendix K to 10 CFR part 50--Emergency Core Cooling System
Evaluation Models:
a. Section I.A.4--Heat generation rates from radioactive decay
of fission products;
b. Section I.A.5--Rate of energy release, hydrogen generation,
and cladding oxidation from the metal/water reaction;
c. Section I.B--Predicting cladding swelling and rupture;
d. Section I.C.1.b--Calculation of the discharge rate for all
times after the discharging fluid has been calculated to be two-
phase;
e. Section I.C.5.a--Post-critical heat flux correlations of heat
transfer from the fuel cladding to the surrounding fluid; and
f. Section I.C.7.a--Calculation of cross-flow between the hot
and average channel regions of the core during blowdown.
VI. Issue Resolution
A. The Commission has determined that the structures, systems,
and components and design features of NuScale comply with the
provisions of the Atomic Energy Act of 1954, as amended, and the
applicable regulations identified in Section V of this appendix; and
therefore, provide adequate protection to the health and safety of
the public. A conclusion that a matter is resolved includes the
finding that additional or alternative structures, systems, and
components, design features, design criteria, testing, analyses,
acceptance criteria, or justifications are not necessary for
NuScale.
B. The Commission considers the following matters resolved
within the meaning of Sec. 52.63(a)(5) in subsequent proceedings
for issuance of a COL, amendment of a COL, or renewal of a COL,
proceedings held under Sec. 52.103, and enforcement proceedings
involving plants referencing this appendix:
1. All nuclear safety issues associated with the information in
the final safety evaluation report, Tier 1, Tier 2, and the
rulemaking record for certification of the NuScale design, with the
exception of the following:
a. generic TS and other operational requirements;
b. the adequacy of the design of the shield wall between the
NuScale power module and the reactor building steam gallery to limit
potential radiological doses consistent with the radiation zones
specified in design certification application Part 2, Tier 2,
Chapter 12, Figure 12.3-1, ``Reactor Building Radiation Zone Map'';
c. the adequacy of the design of the systems used for post-
accident hydrogen and oxygen monitoring described in design
certification application Part 2, Tier 2, Section 6.2.5 to meet the
requirements of 10 CFR 50.34(f)(2)(vii), 10 CFR 50.34(f)(2)(xxviii),
and 10 CFR 52.47(a)(2)(iv), with respect to radiological releases
caused by leakage from these systems under accident conditions; and
d. the ability of the steam generator tubes to maintain
structural and leakage integrity during density wave oscillations in
the secondary fluid system, including the method of analysis to
predict the thermal-hydraulic conditions of the steam generator
secondary fluid system and resulting loads, stresses, and
deformations from density wave oscillations and reverse flow,
consistent with the other design information regarding steam
generator integrity described in DCA Part 2, Tier 2, Sections 3.9.1,
3.9.2, 5.4.1, and 15.6.3, and in accordance with 10 CFR part 50, GDC
4 and 31;
2. All nuclear safety and safeguards issues associated with the
referenced information in the non-public documents in Tables 1.6-1
and 1.6-2 of Tier 2 of the DCD, which contain sensitive unclassified
non-safeguards information (including proprietary information and
security-related information) and safeguards information and which,
in context, are intended as requirements in the generic DCD for the
NuScale design;
3. All generic changes to the DCD under and in compliance with
the change processes in paragraphs VIII.A.1 and VIII.B.1 of this
appendix;
4. All exemptions from the DCD under and in compliance with the
change processes in paragraphs VIII.A.4 and VIII.B.4 of this
appendix, but only for that plant;
5. All departures from the DCD that are approved by license
amendment, but only for that plant;
6. Except as provided in paragraph VIII.B.5.g of this appendix,
all departures from Tier 2 under and in compliance with the change
processes in paragraph VIII.B.5 of this appendix that do not require
prior NRC approval, but only for that plant; and
7. All environmental issues concerning severe accident
mitigation design alternatives associated with the information in
the NRC's environmental assessment for NuScale (ADAMS Accession No.
ML22004A006) and DCD Part 3, ``Applicant's Environmental Report--
Standard Design Certification,'' Revision 5, dated July 2020 (ADAMS
Accession No. ML20224A512), for plants referencing this appendix
whose site characteristics fall within the site parameters of the
representative site specified in the NuScale environmental report.
C. The Commission does not consider operational requirements for
an applicant or
[[Page 3310]]
licensee who references this appendix to be matters resolved within
the meaning of Sec. 52.63(a)(5). The Commission reserves the right
to require operational requirements for an applicant or licensee who
references this appendix by rule, regulation, order, or license
condition.
D. Except under the change processes in Section VIII of this
appendix, the Commission may not require an applicant or licensee
who references this appendix to:
1. Modify structures, systems, and components or design features
as described in the generic DCD;
2. Provide additional or alternative structures, systems, and
components or design features not discussed in the generic DCD; or
3. Provide additional or alternative design criteria, testing,
analyses, acceptance criteria, or justification for structures,
systems, and components or design features discussed in the generic
DCD.
E. The NRC will specify, at an appropriate time, the procedures
to be used by an interested person who wishes to review portions of
the design certification or references containing safeguards
information or sensitive unclassified non-safeguards information
(including proprietary information, such as trade secrets and
commercial or financial information obtained from a person that are
privileged or confidential (10 CFR 2.390 and 10 CFR part 9), and
security-related information), for the purpose of participating in
the hearing required by Sec. 52.85, the hearing provided under
Sec. 52.103, or in any other proceeding relating to this appendix,
in which interested persons have a right to request an adjudicatory
hearing.
VII. Duration of This Appendix
This appendix may be referenced for a period of 15 years from
February 21, 2023, except as provided for in Sec. Sec. 52.55(b) and
52.57(b). This appendix remains valid for an applicant or licensee
who references this appendix until the application is withdrawn or
the license expires, including any period of extended operation
under a renewed license.
VIII. Processes for Changes and Departures
A. Tier 1 Information
1. Generic changes to Tier 1 information are governed by the
requirements in Sec. 52.63(a)(1).
2. Generic changes to Tier 1 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs A.3 or A.4 of this section.
3. Departures from Tier 1 information that are required by the
Commission through plant-specific orders are governed by the
requirements in Sec. 52.63(a)(4).
4. Exemptions from Tier 1 information are governed by the
requirements in Sec. Sec. 52.63(b)(1) and 52.98(f). The Commission
will deny a request for an exemption from Tier 1, if it finds that
the design change will result in a significant decrease in the level
of safety otherwise provided by the design.
B. Tier 2 Information
1. Generic changes to Tier 2 information are governed by the
requirements in Sec. 52.63(a)(1).
2. Generic changes to Tier 2 information are applicable to all
applicants or licensees who reference this appendix, except those
for which the change has been rendered technically irrelevant by
action taken under paragraphs B.3, B.4, or B.5, of this section.
3. The Commission may not require new requirements on Tier 2
information by plant-specific order, while this appendix is in
effect under Sec. 52.55 or Sec. 52.61, unless:
a. A modification is necessary to secure compliance with the
Commission's regulations applicable and in effect at the time this
appendix was approved, as set forth in Section V of this appendix,
or to ensure adequate protection of the public health and safety or
the common defense and security; and
b. Special circumstances as defined in 10 CFR 50.12(a) are
present.
4. An applicant or licensee who references this appendix may
request an exemption from Tier 2 information. The Commission may
grant such a request only if it determines that the exemption will
comply with the requirements of 10 CFR 50.12(a). The Commission will
deny a request for an exemption from Tier 2, if it finds that the
design change will result in a significant decrease in the level of
safety otherwise provided by the design. The granting of an
exemption to an applicant must be subject to litigation in the same
manner as other issues material to the license hearing. The granting
of an exemption to a licensee must be subject to an opportunity for
a hearing in the same manner as license amendments.
5.a. An applicant or licensee who references this appendix may
depart from Tier 2 information, without prior NRC approval, unless
the proposed departure involves a change to or departure from Tier 1
information, or the TS, or requires a license amendment under
paragraph B.5.b or B.5.c of this section. When evaluating the
proposed departure, an applicant or licensee shall consider all
matters described in the plant-specific DCD.
b. A proposed departure from Tier 2, other than one affecting
resolution of a severe accident issue identified in the plant-
specific DCD or one affecting information required by Sec.
52.47(a)(28) to address aircraft impacts, requires a license
amendment if it would:
(1) Result in more than a minimal increase in the frequency of
occurrence of an accident previously evaluated in the plant-specific
DCD;
(2) Result in more than a minimal increase in the likelihood of
occurrence of a malfunction of a structure, system, or component
important to safety and previously evaluated in the plant-specific
DCD;
(3) Result in more than a minimal increase in the consequences
of an accident previously evaluated in the plant-specific DCD;
(4) Result in more than a minimal increase in the consequences
of a malfunction of a structure, system, or component important to
safety previously evaluated in the plant-specific DCD;
(5) Create a possibility for an accident of a different type
than any evaluated previously in the plant-specific DCD;
(6) Create a possibility for a malfunction of a structure,
system, or component important to safety with a different result
than any evaluated previously in the plant-specific DCD;
(7) Result in a design-basis limit for a fission product barrier
as described in the plant-specific DCD being exceeded or altered; or
(8) Result in a departure from a method of evaluation described
in the plant-specific DCD used in establishing the design bases or
in the safety analyses.
c. A proposed departure from Tier 2, affecting resolution of an
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
(1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe
accident previously reviewed and determined to be not credible could
become credible; or
(2) There is a substantial increase in the consequences to the
public of a particular ex-vessel severe accident previously
reviewed.
d. A proposed departure from Tier 2 information required by
Sec. 52.47(a)(28) to address aircraft impacts shall consider the
effect of the changed design feature or functional capability on the
original aircraft impact assessment required by 10 CFR 50.150(a).
The applicant or licensee shall describe, in the plant-specific DCD,
how the modified design features and functional capabilities
continue to meet the aircraft impact assessment requirements in 10
CFR 50.150(a)(1).
e. If a departure requires a license amendment under paragraph
B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
f. A departure from Tier 2 information that is made under
paragraph B.5 of this section does not require an exemption from
this appendix.
g. A party to an adjudicatory proceeding for either the
issuance, amendment, or renewal of a license or for operation under
Sec. 52.103(a), who believes that an applicant or licensee who
references this appendix has not complied with paragraph VIII.B.5 of
this appendix when departing from Tier 2 information, may petition
to admit into the proceeding such a contention. In addition to
complying with the general requirements of 10 CFR 2.309, the
petition must demonstrate that the departure does not comply with
paragraph VIII.B.5 of this appendix. Further, the petition must
demonstrate that the change bears on an asserted noncompliance with
an ITAAC acceptance criterion in the case of a Sec. 52.103
preoperational hearing, or that the departure bears directly on the
amendment request in the case of a hearing on a license amendment.
Any other party may file a response. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall
certify the matter directly to the Commission for determination of
the admissibility of the contention. The Commission may admit such
[[Page 3311]]
a contention if it determines the petition raises a genuine issue of
material fact regarding compliance with paragraph VIII.B.5 of this
appendix.
C. Operational Requirements
1. Changes to NuScale design certification generic TS and other
operational requirements that were completely reviewed and approved
in the design certification rule and do not require a change to a
design feature in the generic DCD are governed by the requirements
in 10 CFR 50.109. Changes that require a change to a design feature
in the generic DCD are governed by the requirements in paragraphs A
or B of this section.
2. Changes to NuScale design certification generic TS and other
operational requirements are applicable to all applicants who
reference this appendix, except those for which the change has been
rendered technically irrelevant by action taken under paragraphs C.3
or C.4 of this section.
3. The Commission may require plant-specific departures on
generic TS and other operational requirements that were completely
reviewed and approved, provided a change to a design feature in the
generic DCD is not required and special circumstances, as defined in
10 CFR 2.335 are present. The Commission may modify or supplement
generic TS and other operational requirements that were not
completely reviewed and approved or require additional TS and other
operational requirements on a plant-specific basis, provided a
change to a design feature in the generic DCD is not required.
4. An applicant who references this appendix may request an
exemption from the generic TS or other operational requirements. The
Commission may grant such a request only if it determines that the
exemption will comply with the requirements of Sec. 52.7. The
granting of an exemption must be subject to litigation in the same
manner as other issues material to the license hearing.
5. A party to an adjudicatory proceeding for the issuance,
amendment, or renewal of a license, or for operation under Sec.
52.103(a), who believes that an operational requirement approved in
the DCD or a TS derived from the generic TS must be changed, may
petition to admit such a contention into the proceeding. The
petition must comply with the general requirements of Sec. 2.309 of
this chapter and must either demonstrate why special circumstances
as defined in Sec. 2.335 of this chapter are present or demonstrate
that the proposed change is necessary for compliance with the
Commission's regulations in effect at the time this appendix was
approved, as set forth in Section V of this appendix. Any other
party may file a response to the petition. If, on the basis of the
petition and any response, the presiding officer determines that a
sufficient showing has been made, the presiding officer shall
certify the matter directly to the Commission for determination of
the admissibility of the contention. All other issues with respect
to the plant-specific TS or other operational requirements are
subject to a hearing as part of the licensing proceeding.
6. After issuance of a license, the generic TS have no further
effect on the plant-specific TS. Changes to the plant-specific TS
will be treated as license amendments under 10 CFR 50.90.
IX. [Reserved]
X. Records and Reporting
A. Records
1. The applicant for this appendix shall maintain a copy of the
generic DCD that includes all generic changes that are made to Tier
1 and Tier 2, and the generic TS and other operational requirements.
The applicant shall maintain the sensitive unclassified non-
safeguards information (including proprietary information and
security-related information) and safeguards information referenced
in the generic DCD for the period that this appendix may be
referenced, as specified in Section VII of this appendix.
2. An applicant or licensee who references this appendix shall
maintain the plant-specific DCD to accurately reflect both generic
changes to the generic DCD and plant-specific departures made under
Section VIII of this appendix throughout the period of application
and for the term of the license (including any periods of renewal).
3. An applicant or licensee who references this appendix shall
prepare and maintain written evaluations that provide the bases for
the determinations required by Section VIII of this appendix. These
evaluations must be retained throughout the period of application
and for the term of the license (including any periods of renewal).
4.a. The applicant for NuScale shall maintain a copy of the
aircraft impact assessment performed to comply with the requirements
of 10 CFR 50.150(a) for the term of the certification (including any
period of renewal).
b. An applicant or licensee who references this appendix shall
maintain a copy of the aircraft impact assessment performed to
comply with the requirements of 10 CFR 50.150(a) throughout the
pendency of the application and for the term of the license
(including any periods of renewal).
B. Reporting
1. An applicant or licensee who references this appendix shall
submit a report to the NRC containing a brief description of any
plant-specific departures from the DCD, including a summary of the
evaluation of each departure. This report must be filed in
accordance with the filing requirements applicable to reports in
Sec. 52.3.
2. An applicant or licensee who references this appendix shall
submit updates to its plant-specific DCD, which reflect the generic
changes to and plant-specific departures from the generic DCD made
under Section VIII of this appendix. These updates shall be filed
under the filing requirements applicable to final safety analysis
report updates in 10 CFR 50.71(e) and 52.3.
3. The reports and updates required by paragraphs X.B.1 and
X.B.2 of this appendix must be submitted as follows:
a. On the date that an application for a license referencing
this appendix is submitted, the application must include the report
and any updates to the generic DCD.
b. During the interval from the date of application for a
license to the date the Commission makes its finding required by
Sec. 52.103(g), the report must be submitted semiannually. Updates
to the plant-specific DCD must be submitted annually and may be
submitted along with amendments to the application.
c. After the Commission makes the finding required by Sec.
52.103(g), the reports and updates to the plant-specific DCD must be
submitted, along with updates to the site-specific portion of the
final safety analysis report for the facility, at the intervals
required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at
shorter intervals as specified in the license.
Dated: January 11, 2023.
For the Nuclear Regulatory Commission.
Brooke P. Clark,
Secretary of the Commission.
[FR Doc. 2023-00729 Filed 1-18-23; 8:45 am]
BILLING CODE 7590-01-P