NuScale Small Modular Reactor Design Certification
| Citation | 86 FR 34999 |
| Record Number | 2021-13940 |
| Published date | 01 July 2021 |
| Court | Nuclear Regulatory Commission |
Federal Register, Volume 86 Issue 124 (Thursday, July 1, 2021)
[Federal Register Volume 86, Number 124 (Thursday, July 1, 2021)]
[Proposed Rules]
[Pages 34999-35023]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2021-13940]
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Proposed Rules
Federal Register
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This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 86, No. 124 / Thursday, July 1, 2021 /
Proposed Rules
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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: Proposed rule.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to
amend 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. The public is invited to
submit comments on this proposed rule.
DATES: Submit comments by August 30, 2021. Comments received after this
date will be considered if it is practical to do so, but the NRC is
able to ensure consideration only for comments received before this
date.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject); however, the NRC encourages electronic
comment submission through the Federal Rulemaking website:
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
individual(s) listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Email comments to: [email protected]. If you do
not receive an automatic email reply confirming receipt, then contact
us at 301-415-1677.
For additional direction on obtaining information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Yanely Malave, Office of Nuclear
Material Safety and Safeguards, telephone: 301-415-1519, email:
[email protected], and Prosanta Chowdhury, Office of Nuclear
Reactor Regulation, telephone: 301-415-1647, email:
[email protected]. Both are staff of the U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001.
SUPPLEMENTARY INFORMATION:
Table of Contents
I. Obtaining Information and Submitting Comments
II. Background
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. Section-by-Section Analysis
VII. Regulatory Flexibility Certification
VIII. Regulatory Analysis
IX. Backfitting and Issue Finality
X. Plain Writing
XI. Environmental Assessment and Finding of No Significant Impact
XII. Paperwork Reduction Act
XIII. Agreement State Compatibility
XIV. Voluntary Consensus Standards
XV. Availability of Documents
XVI. Procedures for Access to Proprietary and Safeguards Information
for Preparation of Comments on the NuScale Design Certification
Proposed Rule
XVII. Incorporation by Reference--Reasonable Availability to
Interested Parties
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2017-0029 when contacting the NRC
about the availability of information for this proposed rule. You may
obtain publicly available information related to this proposed rule by
any of the following methods:
Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0029.
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, at 301-415-4737,
or by email to [email protected]. The ADAMS accession number for
each document referenced in this proposed rule (if that document is
available in ADAMS) is provided the first time that it is mentioned in
this document. In addition, for the convenience of the reader,
instructions about obtaining materials referenced in this document are
provided in Section XV, ``Availability of Documents,'' of this
document.
Attention: The Public Document Room (PDR), where you may
examine and order copies of public documents, is currently closed. You
may submit your request to the PDR via email at [email protected] or
by calling 1-800-397-4209 between 8:00 a.m. and 4:00 p.m. (ET), Monday
through Friday, except Federal holidays.
Attention: 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.
B. Submitting Comments
The NRC encourages electronic comment submission through the
Federal Rulemaking website (https://www.regulations.gov). Please
include Docket ID NRC-2017-0029 in your comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly
[[Page 35000]]
disclosed in your comment submission. The NRC will post all comment
submissions at https://www.regulations.gov as well as enter the comment
submissions into ADAMS. The NRC does not routinely edit comment
submissions to remove identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
II. 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) (ADAMS Accession No. ML17013A229). 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 total capacity of 600 MWe for 12
power modules. All NuScale power modules are partially submerged in one
safety-related pool, which is also the ultimate heat sink for the
reactor. 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.
III. Regulatory and Policy Issues
A. Control Room Staffing Requirements
The requirements in Sec. 50.54(k) and Sec. 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 (ADAMS Accession
No. ML15258A846), 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 (i.e., 12 licensed operators). Absent alternative
staffing requirements, future applicants referencing the NuScale design
would need to request an exemption.
In the DCA Part 7, Section 6.2, ``Justification for Rulemaking,''
NuScale Power provided a technical basis for rulemaking language that
would address control room staffing in conjunction with control room
configuration. NuScale Power's approach is consistent with SECY-11-
0098, ``Operator Staffing for Small or Multi-Module Nuclear Power Plant
Facilities,'' dated July 22, 2011 (ADAMS Accession No. ML111870574). In
Chapter 18, Section 18.5.4.2, ``Evaluation of the Applicant's Technical
Basis,'' of the final safety evaluation report (ADAMS Accession No.
ML20023B605), 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 proposed rule
includes the alternative staffing requirement provisions, staffing
table, and appropriate table notes, a future applicant or licensee that
references proposed appendix G to 10 CFR part 52 would not need to
request an exemption from Sec. 50.54(m).
B. Incorporation by Reference
The proposed Section III.A, ``Incorporation by reference
approval,'' of appendix G to 10 CFR part 52 lists documents that would
be approved by the Director of the Office of the Federal Register for
incorporation by reference into this appendix. Proposed 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.''
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
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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. A COL applicant would be 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. A COL 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 a COL applicant that references this appendix without a
demonstrable impact on safety or standardization. Appendix G to 10 CFR
part 52, Section VI, ``Issue Resolution,'' would clarify that this
issue is not resolved within the meaning of Sec. 52.63(a)(5), and
Section IV, ``Additional Requirements and Restrictions,'' would state
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 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.
A COL applicant would be required 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 COL 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 would likely 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 COL application
review to provide reasonable assurance of adequate protection.
Therefore, the NRC has determined that this information can be provided
by a COL applicant that references this appendix without a demonstrable
impact on safety or standardization. Appendix G to 10 CFR part 52,
Section VI, ``Issue Resolution,'' would clarify that this issue is not
resolved within the meaning of Sec. 52.63(a)(5), and Section IV,
``Additional Requirements and Restrictions,'' would state 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, 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 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 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 nuclear 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 part 20 and 10 CFR part 50, appendix A, General Design Criterion
(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,
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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, an appropriate method of
analysis has not been provided to the NRC.
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 TR-1016-51669 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 acknowledges 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, ``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 would 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 part of the NuScale DCA but would be
accomplished 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 a COL 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,'' would clarify that
this issue is not resolved within the meaning of Sec. 52.63(a)(5), and
Section IV, ``Additional Requirements and Restrictions,'' would state
that the COL applicant is responsible for providing the design
information to address this issue.
IV. Technical Issues Associated With the NuScale Design
The NRC identified significant technical issues associated with the
following design areas that were resolved by NuScale Power 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'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, Components,
Equipment, and Systems,'' 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,
[[Page 35003]]
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 previously, 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 which result in
transient thermal-hydraulic responses that are different from those of
currently licensed reactors.
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-P, Revision 3, ``Containment
Response Analysis Methodology,'' May 2020 (ADAMS Accession No.
ML20141L808) that describes the conservative containment pressure and
temperature safety analyses for several design-basis events related to
the containment design margins. NuScale also submitted topical report
TR-0516-49422, ``Loss-of-Coolant Accident Evaluation Model,'' Revision
1, dated November 2019 (ADAMS Accession No. ML19331B585). 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 (ADAMS Accession No. ML20044E199). 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 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 satisfy 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, Components, Equipment, and Systems,'' 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 does not apply the single failure criterion to the IAB valve,
and 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,'' (ADAMS
Accession No. ML19183A408) 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
[[Page 35004]]
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'' (ADAMS Accession No.
ML060260236), 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'' (ADAMS
Accession No. ML003708098), 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 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. Exemption to 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''' (ADAMS Accession No. ML18065A431), dated October 9, 2018.
The NRC granted the exemption request.
E. 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, ``Safety
Classification of Passive Nuclear Power Plant Electrical Systems,''
dated February 23, 2018 (ADAMS Accession No. ML18054B607). 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 (ADAMS
Accession No. ML17048A459) issued December 13, 2017, as described in
the final safety evaluation report for Chapter 1, ``Introduction and
General Discussion,'' (ADAMS Accession No. ML20204A986).
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'' (ADAMS
Accession No. ML20023B614).
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 which have been used in
safety and siting assessments for large light water reactors. The
design-basis accidents for
[[Page 35005]]
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 (ADAMS Accession No. ML19112A172).
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.
The 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 (ADAMS Accession No. ML19107A455), describing the regulatory and
technical issues raised by unique aspects of NuScale Power's proposed
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 29, 2019 (ADAMS Accession No.
ML19297G520). The approved version TR-0915-17565-NP-A, Revision 4
(ADAMS Accession No. ML20057G132) is discussed in the DCA 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,'' 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, ``Exemption to 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 (ADAMS Accession No. ML20204A986)):
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''
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 reports where these exemption requests have been evaluated.
The NRC granted each exemption request.
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 (ADAMS Accession
No. ML20023A318) after the Advisory Committee on Reactor Safeguards
(ACRS) performed its final independent review and issued its letter to
the Commission in July 2020 on its findings
[[Page 35006]]
and recommendations (ADAMS Accession No. ML20211M386). 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 reflects that NuScale Power has resolved
all technical and safety issues with the exception of the three issues
discussed previously. The final safety evaluation report describes the
portions of the design that are not receiving finality in this rule
and, therefore, will not be part of the certified design. The final
safety evaluation report 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 Proposed Rule
The following discussion describes the purpose and key aspects of
each section of this NuScale design certification proposed 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 proposed 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 would be approved by this design
certification proposed 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 for a COL contracts with the design certification applicant
to obtain the generic DCD and supporting design information. If the COL
applicant does not use the design certification applicant to provide
the design information and instead uses an alternate nuclear plant
vendor, then the COL applicant must meet the requirements in Sec.
52.73. Second, paragraph X.A.1 would require 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
proposed 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 will be 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 may also have to
include considerations for multi-module facilities in the plant-
specific DCD that were not previously evaluated as part of the design
certification rule, depending on the contents of the application.
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 Chapter 16,
``Technical Specifications,'' of the generic DCD 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, dated February 15, 1991, on SECY-90-377, ``Requirements for
Design Certification under 10 CFR part 52,'' dated November 8, 1990
(ADAMS Accession No. ML003707892).
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
[[Page 35007]]
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 proposed 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 COL applicants and licensees
referencing the NuScale DCD. 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
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.
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 would be resolved by the NRC through
this proposed 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 would be resolved by this proposed
rule, paragraph VI.C identifies issues which are not resolved by this
proposed 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 proposed 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
[[Page 35008]]
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 proposed 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.
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\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 for a COL, 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 COL 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.
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 would require 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 (ADAMS Accession No.
ML16196A321), 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
[[Page 35009]]
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 has used 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 would provide 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 may
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) 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 which 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 would be permitted to 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 may 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
[[Page 35010]]
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 would allow 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 would be
similar to the requirement in Sec. 50.59 and would apply 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 requirements of 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 in Chapter 16 of the DCD or other
operational requirements in the generic DCD would be 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 would be 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
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 the applicant for a COL
referencing the NuScale design certification rule 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 (ADAMS Accession No.
ML18131A181)). 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 specifications 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 would be
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-
[[Page 35011]]
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 proposed 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
this design certification rule. Only the generic DCD is identified and
incorporated by reference into this 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 would be incorporated 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 proposed rule does not require that
duplicate documentation be maintained by an applicant or licensee that
references this 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 will be 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.
[[Page 35012]]
VI. Section-by-Section Analysis
The following paragraphs describe the specific changes of this
proposed rule:
Section 52.11, Information collection requirements: Office of
Management and Budget (OMB) approval.
In Sec. 52.11, this proposed rule would add 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 proposed rule would add 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.
VII. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC
certifies that this rule, if promulgated, will not have a significant
economic impact on a substantial number of small entities. This
proposed 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).
VIII. Regulatory Analysis
The NRC has not prepared a regulatory analysis for this proposed
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.
IX. Backfitting and Issue Finality
The NRC has determined that this proposed 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 proposed 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 proposed 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 proposed rule.
X. 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 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). The NRC requests comment on the proposed rule
with respect to clarity and effectiveness of the language used.
XI. Environmental Assessment and Finding of No Significant Impact
The NRC conducted an environmental assessment (ADAMS Accession No.
ML19303C179) 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 proposed 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. Comments on the environmental assessment will
be limited to the consideration of severe accident mitigation design
alternatives as required by Sec. 51.30(d).
The basis for the NRC's categorical exclusion in this regard, as
discussed in the 2007 final rule amending 10 CFR parts 51 and 52 (72 FR
49352; 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 DC rule.
Consistent with Sec. 51.30(d) and Sec. 51.32(b), the NRC has
prepared an environmental assessment (ADAMS Accession No. ML19303C179)
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'' (ADAMS Accession No.
ML20224A512). 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 those site parameters 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 given site parameters. Site-specific SAMDAs,
[[Page 35013]]
multi-unit aspects, procedural and training SAMDAs, and the reactor
building crane design would need to be assessed when a specific site is
proposed for constructing and operating 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 XV
of this proposed rule.
XII. Paperwork Reduction Act
This proposed rule contains new or amended collections of
information subject to the Paperwork Reduction Act of 1995 (44 U.S.C.
3501 et seq). This proposed rule has been submitted to the OMB for
review and approval of the information collections.
Type of submission: Revision.
The title of the information collection: Appendix G to 10 CFR part
52 Design Certification Rule for NuScale.
The form number if applicable: NA.
How often the collection is required or requested: On occasion
Who will be required or asked to respond: Applicant for a combined
license, construction permit, or a design certification amendment.
An estimate of the number of annual responses: 5 (2 annual
responses and 3 recordkeepers).
The estimated number of annual respondents: 3.
An estimate of the total number of hours needed annually to comply
with the information collection requirement or request: 389 hours (346
reporting hours + 43 recordkeeping hours).
Abstract: The NRC is proposing to amend its regulations to certify
the NuScale standard design. This action is necessary so that
applicants or licensees intending to construct and operate an 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.
The NRC is seeking public comment on the potential impact of the
information collection contained in this proposed rule and on the
following issues:
(1) Is the proposed information collection necessary for the proper
performance of the functions of the NRC, including whether the
information will have practical utility?
(2) Is the estimate of the burden of the proposed information
collection accurate?
(3) Is there a way to enhance the quality, utility, and clarity of
the information to be collected?
(4) How can the burden of the proposed information collection on
respondents be minimized, including the use of automated collection
techniques or other forms of information technology?
A copy of the OMB clearance package is available in ADAMS under
Accession No. ML20242A000 or can be obtained free of charge by
contacting the NRC's Public Document Room reference staff at 1-800-397-
4209, at 301-415-4737, or by email to [email protected] You may
obtain information and comment submissions related to the OMB clearance
package by searching on https://www.regulations.gov under Docket ID
NRC-2017-0029.
You may submit comments on any aspect of these proposed information
collection(s), including suggestions for reducing the burden and on the
above issues, by the following methods:
Federal Rulemaking Website: Go to https://www.regulations.gov and 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].
Additionally, this proposed rule provides procedures for requesting
access to proprietary and safeguards information for preparation of
comments on the NuScale design certification proposed rule. These
procedures are guidance for completing mandatory information
collections located in 10 CFR parts 9 and 73 that are subject to the
Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These
information collections were approved by OMB under approval numbers
3150-0043 and 3150-0002. Send comments regarding this information
collection to the FOIA, Library, and Information Collections Branch
(T6-A10M), U.S. Nuclear Regulatory Commission, Washington, DC 20555
0001, or by email to [email protected], and to the OMB
reviewer at: OMB Office of Information and Regulatory Affairs (3150-
0043 and 3150-0002), Attn: Desk Officer for the Nuclear Regulatory
Commission, 725 17th Street NW, Washington, DC 20503; email:
[email protected].
Submit comments by August 30, 2021. Comments received after this
date will be considered if it is practical to do so, but the NRC is
able to ensure consideration only for comments received on or before
this date.
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.
XIII. Agreement State Compatibility
Under the ``Policy Statement on Adequacy and Compatibility of
Agreement States Programs,'' approved by the Commission on June 20,
1997, and published in the Federal Register (62 FR 46517; September 3,
1997), this proposed 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 Atomic Energy Act or the
provisions of 10 CFR, and although an Agreement State may not adopt
program elements reserved to the NRC, it may wish to inform its
licensees of certain requirements by a mechanism that is consistent
with a particular State's administrative procedure laws, but does not
confer regulatory authority on the State.
XIV. 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 proposed rule, the NRC
intends to certify 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.
XV. Availability of Documents
The documents identified in the following table are available to
[[Page 35014]]
interested persons through one or more of the following methods, as
indicated.
------------------------------------------------------------------------
ADAMS
Document accession No.
------------------------------------------------------------------------
SECY-21-0004, ``Proposed Rule: NuScale Small Modular ML19353A003
Reactor Design Certification (RIN 3150-AJ98; NRC-2017-
0029)''................................................
Staff Requirements Memorandum for SECY-21-0004, ML21126A153
``Proposed Rule: NuScale Small Modular Reactor Design
Certification (RIN 3150-AJ98; NRC-2017-0029)''.........
NuScale Power, LLC Submittal of the NuScale Standard ML17013A229
Plant Design Certification Application (NRC Project No.
0769) (December 2016)..................................
NuScale Power, LLC Submittal of the NuScale Standard ML20225A071
Plant Design Certification Application, Revision 5
(July 2020)............................................
NuScale DCA Final Safety Evaluation Reports (August ML20023A318
2020)..................................................
NuScale Standard Design Certification Application, Part ML20224A512
3, ``Applicant's Environmental Report--Standard Design
Certification,'' Revision 5 (July 2020)................
Environmental Assessment by the U.S. Nuclear Regulatory ML19303C179
Commission Relating to the Certification of the NuScale
Standard Design........................................
Regulatory History of Design Certification (April 2000) ML003761550
\2\....................................................
------------------------------------------------------------------------
NuScale Technical and Topical Reports
------------------------------------------------------------------------
ES-0304-1381-NP, Human-System Interface Style Guide, ML19338E948
Rev. 4 (December 2019).................................
RP-0215-10815-NP, Concept of Operations, Rev. 3 (May ML19133A293
2019)..................................................
RP-0316-17614-NP, Human Factors Engineering Operating ML16364A342
Experience Review Results Summary Report, Rev. 0
(December 2016)........................................
RP-0316-17615-NP, Human Factors Engineering Functional ML16364A342
Requirements Analysis and Function Allocation Results
Summary Report, Rev. 0 (December 2016).................
RP-0316-17616-NP, Human Factors Engineering Task ML19119A393
Analysis Results Summary Report, Rev. 2 (April 2019)...
RP-0316-17617-NP, Human Factors Engineering Staffing and ML17004A222
Qualifications Results Summary Report, Rev. 0 (December
2016)..................................................
RP-0316-17618-NP, Human Factors Engineering Treatment of ML17004A222
Important Human Actions Results Summary Report, Rev. 0
(December 2016)........................................
RP-0316-17619-NP, Human Factors Engineering Human-System ML19119A398
Interface Design Results Summary Report, Rev. 2, (April
2019)..................................................
RP-0516-49116-NP, Control Room Staffing Plan Validation ML16364A356
Results, Rev. 1 (December 2016)........................
RP-0914-8534-NP, Human Factors Engineering Program ML19119A342
Management Plan, Rev. 5 (April 2019)...................
RP-0914-8543-NP, Human Factors Verification and ML19119A372
Validation Implementation Plan, Rev. 5 (April 2019)....
RP-0914-8544-NP, Human Factors Engineering Design ML19331A910
Implementation Implementation Plan, Rev. 4 (November
2019)..................................................
RP-1018-61289-NP, Human Factors Engineering Verification ML19212A773
and Validation Results Summary Report, Rev. 1 (July
2019)..................................................
RP-1215-20253-NP, Control Room Staffing Plan Validation ML16364A353
Methodology, Rev. 3 (December 2016)....................
TR-0116-20781-NP, Fluence Calculation Methodology and ML19183A485
Results, Rev. 1 (July 2019)............................
TR-0116-20825-NP-A, Applicability of AREVA Fuel ML18040B306
Methodology for the NuScale Design, Rev. 1 (February
2018)..................................................
TR-0116-21012-NP-A, NuScale Power Critical Heat Flux ML18360A632
Correlations, Rev. 1 (December 2018)...................
TR-0316-22048-NP, Nuclear Steam Supply System Advanced ML20141M764
Sensor Technical Report, Rev. 3 (May 2020).............
TR-0515-13952-NP-A, Risk Significance Determination, ML16284A016
Rev. 0 (October 2016)..................................
TR-0516-49084-NP, Containment Response Analysis ML20141L808
Methodology Technical Report, Rev. 3 (May 2020)........
TR-0516-49416-NP-A, Non-Loss-of-Coolant Accident ML20191A281
Analysis Methodology, Rev. 3 (July 2020)...............
TR-0516-49417-NP-A, Evaluation Methodology for Stability ML20078Q094
Analysis of the NuScale Power Module, Rev. 1 (March
2020)..................................................
TR-0516-49422-NP-A, Loss-of-Coolant Accident Evaluation ML20189A644
Model, Rev. 2 (July 2020)..............................
TR-0616-48793-NP-A, Nuclear Analysis Codes and Methods ML18348B036
Qualification, Rev. 1 (December 2018)..................
TR-0616-49121-NP, NuScale Instrument Setpoint ML20141M114
Methodology Technical Report, Rev. 3 (May 2020)........
TR-0716-50350-NP-A, Rod Ejection Accident Methodology, ML20168B203
Rev. 1 (June 2020).....................................
TR-0716-50351-NP-A, NuScale Applicability of AREVA ML20122A248
Method for the Evaluation of Fuel Assembly Structural
Response to Externally Applied Forces, Rev. 1 (May
2020)..................................................
TR-0716-50424-NP, Combustible Gas Control, Rev. 1 (March ML19091A232
2019)..................................................
TR-0716-50439-NP, NuScale Comprehensive Vibration ML19212A776
Assessment Program Analysis Technical Report, Rev. 2
(July 2019)............................................
TR-0815-16497-NP-A, Safety Classification of Passive ML18054B607
Nuclear Power Plant Electrical Systems Topical Report,
Rev. 1 (February 2018).................................
TR-0816-49833-NP, Fuel Storage Rack Analysis, Rev. 1 ML18310A154
(November 2018)........................................
TR-0816-50796-NP, Loss of Large Areas Due to Explosions ML19165A294
and Fires Assessment, Rev. 1 (June 2019)...............
TR-0816-50797 (NuScale Nonproprietary), Mitigation ML19302H598
Strategies for Loss of All AC Power Event, Rev. 3
(October 2019).........................................
TR-0816-51127-NP, NuFuel-HTP2TM Fuel and Control Rod ML19353A719
Assembly Designs, Rev. 3 (December 2019)...............
TR-0818-61384-NP, Pipe Rupture Hazards Analysis, Rev. 2 ML19212A682
(July 2019)............................................
TR-0915-17564-NP-A, Subchannel Analysis Methodology, ML19067A256
Rev. 2 (March 2019)....................................
TR-0915-17565-NP-A, Accident Source Term Methodology, ML20057G132
Rev. 4 (February 2020).................................
TR-0916-51299-NP, Long-Term Cooling Methodology, Rev. 3 ML20141L816
(May 2020).............................................
TR-0916-51502-NP, NuScale Power Module Seismic Analysis, ML19093B850
Rev. 2 (April 2019)....................................
TR-0917-56119-NP, CNV Ultimate Pressure Integrity, Rev. ML19158A382
1 (June 2019)..........................................
TR-0918-60894-NP, Comprehensive Vibration Assessment ML19214A248
Program Measurement and Inspection Plan Technical
Report, Rev, 1 (August 2019)...........................
TR-1010-859-NP-A, NuScale Topical Report: Quality ML20176A494
Assurance Program Description for the NuScale Power
Plant, Rev. 5 (June 2020)..............................
TR-1015-18177-NP, Pressure and Temperature Limits ML18298A304
Methodology, Rev. 2 (October 2018).....................
TR-1015-18653-NP-A, Design of the Highly Integrated ML17256A892
Protection System Platform Topical Report, Rev. 2
(September 2017).......................................
TR-1016-51669-NP, NuScale Power Module Short-Term ML19211D411
Transient Analysis, Rev. 1 (July 2019).................
TR-1116-51962-NP, NuScale Containment Leakage Integrity ML19149A298
Assurance Technical Report, Rev. 1 (May 2019)..........
[[Page 35015]]
TR-1116-52065-NP, Effluent Release (GALE Replacement) ML18317A364
Methodology and Results, Rev. 1 (November 2018)........
------------------------------------------------------------------------
---------------------------------------------------------------------------
\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.
---------------------------------------------------------------------------
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.
XVI. Procedures for Access to Proprietary and Safeguards Information
for Preparation of Comments on the NuScale Design Certification
Proposed Rule
This section contains instructions regarding how the non-publicly
available documents related to this rule, and specifically those listed
in Table 1.6-1 and 1.6-2 beginning on page 1.6-2 of Tier 2 of the DCD,
may be accessed by interested persons who wish to comment on the design
certification. These documents contain proprietary information and
safeguards information (SGI). Requirements for access to SGI are
primarily set forth in 10 CFR parts 2 and 73. This section provides
information specific to this proposed rule; however, nothing in this
section is intended to conflict with the SGI regulations.
Interested persons who desire access to proprietary information on
NuScale should first request access to that information from NuScale
Power, LLC, the design certification applicant. Requests to the
applicant must be sent to NuScale Power, LLC, at
[email protected]. A request for access should be
submitted to the NRC if the applicant does not either grant or deny
access by the 10-day deadline described in the following section.
One of the non-publicly available documents, TR-0416-48929,
``NuScale Design of Physical Security Systems,'' contains both
proprietary information and SGI. If you need access to proprietary
information in that document in order to develop comments within the
scope of this rule, then your request for access should first be
submitted to NuScale Power, in accordance with the previous paragraph.
By contrast, if you need access to the SGI in order to provide
comments, then your request for access to the SGI must be submitted to
the NRC as described further in this section. Therefore, if you need
access to both proprietary information and SGI in that document, then
you should request access to the information in separate requests
submitted to both NuScale Power and the NRC.
Submitting a Request to the NRC for Access
Within 10 days after publication of this proposed rule, any
individual or entity who believes access to proprietary information or
SGI is necessary in order to submit comments on this proposed rule may
request access to such information. Requests for access to proprietary
information or SGI submitted more than 10 days after publication of
this document will not be considered absent a showing of good cause for
the late filing explaining why the request could not have been filed
earlier.
The requestor shall submit a letter requesting permission to access
proprietary information and/or SGI to the Office of the Secretary, U.S.
Nuclear Regulatory Commission, Attention: Rulemakings and Adjudications
Staff, Washington, DC 20555-0001. The email address for the Office of
the Secretary is [email protected]. The requester must send a
copy of the request to the design certification applicant at the same
time as the original transmission to the NRC using the same method of
transmission. Requests to the applicant must be sent to NuScale Power,
LLC, at [email protected].
The request must include the following information:
(1) The name of this design certification, NuScale Design
Certification; the rulemaking identification number, RIN 3150-AJ98; the
rulemaking docket number, NRC-2017-0029; and the Federal Register
citation for this rule.
(2) The name and address of the requester.
(3) The identity of the individual(s) to whom access is to be
provided, including the identity of any expert, consultant, or
assistant who will aid the requestor in evaluating the information.
(4) If the request is for proprietary information, the requester's
need for the information in order to prepare meaningful comments on the
design certification must be demonstrated. Each of the following areas
must be addressed with specificity:
(a) The specific issue or subject matter on which the requester
wishes to comment.
(b) An explanation why information which is publicly available is
insufficient to provide the basis for developing meaningful comment on
the NuScale design certification proposed rule with respect to the
issue or subject matter described in paragraph 4.a. of this section.
(c) The technical competence (demonstrable knowledge, skill,
training or education) of the requestor to effectively utilize the
requested proprietary information to provide the basis for meaningful
comment. Technical competence may be shown by reliance on a qualified
expert, consultant, or assistant who satisfies these criteria.
(d) A chronology and discussion of the requester's attempts to
obtain the information from the design certification applicant, and the
final communication from the requester to the applicant and the
applicant's response, if any was provided, with respect to the request
for access to proprietary information must be submitted.
(5) If the request is for SGI, the request must include the
following:
(a) A statement that explains each individual's ``need to know''
the SGI, as required by Sec. Sec. 73.2 and 73.22(b)(1). Consistent
with the definition of ``need to know'' as stated in Sec. 73.2, the
statement must explain:
(i) Specifically why the requestor believes that the information is
necessary to enable the requestor to proffer and/or adjudicate a
specific contention in this proceeding; \3\ and
---------------------------------------------------------------------------
\3\ Broad SGI requests under these procedures are unlikely to
meet the standard for need to know. Furthermore, NRC redaction of
information from requested documents before their release may be
appropriate to comport with this requirement. The procedures in this
document do not authorize unrestricted disclosure or less scrutiny
of a requester's need to know than ordinarily would be applied in
connection with either adjudicatory or non-adjudicatory access to
SGI.
---------------------------------------------------------------------------
(ii) The technical competence (demonstrable knowledge, skill,
training or education) of the requestor to effectively utilize the
requested SGI to provide the basis and specificity for meaningful
comment. Technical competence may be shown by reliance
[[Page 35016]]
on a qualified expert, consultant, or assistant who satisfies these
criteria.
(b) A completed Form SF-85, ``Questionnaire for Non-Sensitive
Positions,'' for each individual who would have access to SGI. The
completed Form SF-85 will be used by the Office of Administration to
conduct the background check required for access to SGI, as required by
10 CFR part 2, subpart C, and Sec. 73.22(b)(2), to determine the
requestor's trustworthiness and reliability. For security reasons, Form
SF-85 can be submitted only electronically through the Electronic
Questionnaires for Investigations Processing website, a secure website
that is owned and operated by the Defense Counterintelligence and
Security Agency (DCSA). To obtain online access to the form, the
requestor should contact the NRC's Office of Administration at 301-415-
3710.\4\
---------------------------------------------------------------------------
\4\ The requester will be asked to provide his or her full name,
social security number, date and place of birth, telephone number,
and email address. After providing this information, the requestor
usually should be able to obtain access to the online form within
one business day.
---------------------------------------------------------------------------
(c) A completed Form FD-258 (fingerprint card), signed in original
ink, and submitted in accordance with Sec. 73.57(d). Copies of Form
FD-258 may be obtained by sending an email to [email protected]
or by sending a written request to U.S. Nuclear Regulatory Commission,
Attn: Mailroom/Fingerprint Card Request, 11555 Rockville Pike,
Rockville, MD 20852. The fingerprint card will be used to satisfy the
requirements of 10 CFR part 2, subpart C, Sec. 73.22(b)(1), and
Section 149 of the Atomic Energy Act of 1954, as amended, which
mandates that all persons with access to SGI must be fingerprinted for
an FBI identification and criminal history records check.
(d) A check or money order in the amount of $326.00 \5\ payable to
the U.S. Nuclear Regulatory Commission for each individual for whom the
request for access has been submitted; and
---------------------------------------------------------------------------
\5\ This fee is subject to change pursuant to DCSA's adjustable
billing rates.
---------------------------------------------------------------------------
(e) If the requester or any individual who will have access to SGI
believes they belong to one or more of the categories of individuals
that are exempt from the criminal history records check and background
check requirements, as stated in Sec. 73.59, the requester should also
provide a statement identifying which exemption the requester is
invoking, and explaining the requester's basis for believing that the
exemption applies. While processing the request, the Office of
Administration, Personnel Security Branch, will make a final
determination whether the claimed exemption applies. Alternatively, the
requester may contact the Office of Administration for an evaluation of
their exemption status prior to submitting their request. Persons who
are exempt from the background check are not required to complete the
SF-85 or Form FD-258; however, all other requirements for access to
SGI, including the need to know, are still applicable.
Note: Copies of documents and materials required by paragraphs
(5)(b), (c), and (d), of this section must be sent to the following
address: U.S. Nuclear Regulatory Commission, ATTN: Personnel Security
Branch, Mail Stop TWFN-07D04M, 11555 Rockville Pike, Rockville, MD
20852.
These documents and materials should not be included with the
request letter to the Office of the Secretary, but the request letter
should state that the forms and fees have been submitted as required.
To avoid delays in processing requests for access to SGI, all forms
should be reviewed for completeness and accuracy (including legibility)
before submitting them to the NRC. The NRC will return incomplete or
illegible packages to the sender without processing.
Based on an evaluation of the information submitted under
paragraphs (4) or (5) of this section, as applicable, the NRC will
determine within 10 days of receipt of the request whether the
requester has established a legitimate need for access to proprietary
information or need to know the SGI requested.
Determination of Legitimate Need for Access
For proprietary information access requests, if the NRC determines
that the requester has established a legitimate need for access to
proprietary information, the NRC will notify the requester in writing
that access to proprietary information has been granted. The written
notification will contain instructions on how the requestor may obtain
copies of the requested documents, and any other conditions that may
apply to access to those documents. These conditions may include, but
are not limited to, the signing of a Non-Disclosure Agreement or
Affidavit by each individual who will be granted access.
For requests for access to SGI, if the NRC determines that the
requester has established a need to know the SGI, the NRC's Office of
Administration will then determine, based upon completion of the
background check, whether the proposed recipient is trustworthy and
reliable, as required for access to SGI by Sec. 73.22(b). If the NRC's
Office of Administration determines that the individual or individuals
are trustworthy and reliable, the NRC will promptly notify the
requester in writing. The notification will provide the names of
approved individuals as well as the conditions under which the SGI will
be provided. Those conditions may include, but are not limited to, the
signing of a Non-Disclosure Agreement or Affidavit by each individual
who will be granted access to SGI.
Release and Storage of SGI
Prior to providing SGI to the requester, the NRC will conduct (as
necessary) an inspection to confirm that the recipient's information
protection system is sufficient to satisfy the requirements of Sec.
73.22. Alternatively, recipients may opt to view SGI at an approved SGI
storage location rather than establish their own SGI protection program
to meet SGI protection requirements.
Filing of Comments on the NuScale Design Certification Proposed Rule
Based on Non-Public Information
Any comments in this rulemaking proceeding that are based upon the
information received as a result of the request made for proprietary or
SGI information must be filed by the requester no later than 25 days
after receipt of (or access to) that information, or the close of the
public comment period, whichever is later. The commenter must comply
with all NRC requirements regarding the submission of proprietary
information and SGI to the NRC when submitting comments to the NRC
(including marking and transmission requirements).
Review of Denials of Access
If the request for access to proprietary information or SGI is
denied by the NRC, either after a determination on requisite need or
after a determination on trustworthiness and reliability, the NRC shall
promptly notify the requester in writing, briefly stating the reason or
reasons for the denial.
Before the Office of Administration makes a final adverse
determination regarding the trustworthiness and reliability of the
proposed recipient(s) for access to SGI, the Office of Administration,
in accordance with Sec. 2.336(f)(1)(iii), must provide the proposed
recipient(s) any records that were considered in the trustworthiness
and reliability determination, including those required to be provided
under Sec. 73.57(e)(1), so that the proposed
[[Page 35017]]
recipient(s) have an opportunity to correct or explain the record.
The requestor may challenge the NRC's adverse determination with
respect to access to proprietary information or with respect to need to
know for SGI by filing a challenge within 5 days of receipt of that
determination with the NRC's Executive Director for Operations under
Sec. 9.29(d).
The requestor may challenge the Office of Administration's final
adverse determination with respect to trustworthiness and reliability
for access to SGI by filing a request for review in accordance with
Sec. 2.336(f)(1)(iv).
XVII. Incorporation by Reference--Reasonable Availability to Interested
Parties
The NRC proposes to incorporate 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. The OFR regulations require an agency to include in a proposed rule
a discussion of the ways that the materials the agency incorporates by
reference are reasonably available to interested parties or how it
worked to make those materials reasonably available to interested
parties. The discussion in this section complies with the requirement
for a proposed rule 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 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 \6\ Indian
tribes.
---------------------------------------------------------------------------
\6\ 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
XV of this proposed rule, documents related to this proposed 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 proposed 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 proposes the following
amendments to 10 CFR part 52:
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), add ``G,'' in alphabetical order to the list of
appendices.
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
NuScale, in accordance with 10 CFR part 52, subpart B. The applicant
for the standard design certification of NuScale is NuScale Power,
LLC.
II. Definitions
A. Generic design control document (generic DCD) means the
document containing the Tier 1 and Tier 2 information (including the
technical and topical reports referenced in Chapter 1) and generic
technical specifications that is 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.
[[Page 35018]]
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 G. 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. All other terms in this appendix have the meaning set out in
10 CFR 50.2, 10 CFR 52.1, or Section 11 of the Atomic Energy Act of
1954, as amended, as applicable.
III. Scope and Contents
A. Incorporation by reference approval.
NuScale standard design (hereafter referred as NuScale) material
is approved for incorporation by reference by the Director of the
Office of the Federal Register under 5 U.S.C. 552(a) and 1 CFR part
51, ``Incorporation by Reference.'' You may obtain copies of the
generic DCD from NuScale Power, LLC, 6650 SW Redwood Lane, Suite
210, Portland, Oregon 97224. You can view the generic DCD online in
the NRC Library at https://www.nrc.gov/reading-rm/adams.html. In
ADAMS, search under ADAMS Accession No. ML20225A071. 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]. Copies of the NuScale materials are available
in the ADAMS Public Documents collection. All approved material is
available for inspection at the National Archives and Records
Administration (NARA). For information on the availability of this
material at NARA, email at [email protected] or go to https://www.archives.gov/federal-register/cfr/ibrlocations.html.
1. NuScale Standard Plant Design Certification Application,
Certified Design Descriptions and Inspections, Tests, Analyses, &
Acceptance Criteria (ITAAC), Part 2--Tier 1, Revision 5, July 2020.
2. NuScale Standard Plant Design Certification Application, Part
2--Tier 2, Revision 5, July 2020, including:
a. Chapter One, Introduction and General Description of the
Plant.
b. Chapter Two, Site Characteristics and Site Parameters.
c. Chapter Three, Design of Structures, Systems, Components and
Equipment.
d. Chapter Four, Reactor.
e. Chapter Five, Reactor Coolant System and Connecting Systems.
f. Chapter Six, Engineered Safety Features.
g. Chapter Seven, Instrumentation and Controls.
h. Chapter Eight, Electric Power.
i. Chapter Nine, Auxiliary Systems.
j. Chapter Ten, Steam and Power Conversion System.
k. Chapter Eleven, Radioactive Waste Management.
l. Chapter Twelve, Radiation Protection.
m. Chapter Thirteen, Conduct of Operations.
n. Chapter Fourteen, Initial Test Program and Inspections,
Tests, Analyses, and Acceptance Criteria.
o. Chapter Fifteen, Transient and Accident Analyses.
p. Chapter Sixteen, Technical Specifications.
q. Chapter Seventeen, Quality Assurance and Reliability
Assurance.
r. Chapter Eighteen, Human Factors Engineering.
s. Chapter Nineteen, Probabilistic Risk Assessment and Severe
Accident Evaluation.
t. Chapter Twenty, Mitigation of Beyond-Design-Basis Events.
u. Chapter Twenty-One, Multi-Module Design Considerations.
3. DCA Part 4, Volume 1, Revision 5.0, Generic Technical
Specifications, NuScale Nuclear Power Plants, Volume 1:
Specifications.
4. DCA Part 4, Volume 2, Revision 5.0, Generic Technical
Specifications, NuScale Nuclear Power Plants, Volume 2: Bases.
5. ES-0304-1381-NP, Human-System Interface Style Guide, December
2019, Revision 4, Docket: 52-048.
6. RP-0215-10815-NP, Concept of Operations, May 2019, Revision
3, Docket: 52-048.
7. RP-0316-17614-NP, Human Factors Engineering Operating
Experience Review Results Summary Report, 12/07/2016, Revision 0,
Docket: PROJ0769.
8. RP-0316-17615-NP, Human Factors Engineering Functional
Requirements Analysis and Function Allocation Results Summary
Report, 12/2/16, Revision 0, Docket: PROJ0769.
9. RP-0316-17616-NP, Human Factors Engineering Task Analysis
Results Summary Report, April 2019, Revision 2, Docket: 52-048.
10. RP-0316-17617-NP, Human Factors Engineering Staffing and
Qualifications Results Summary Report, 12/02/2016, Revision 0,
Docket: PROJ0769.
11. RP-0316-17618-NP, Human Factors Engineering Treatment of
Important Human Actions Results Summary Report, 12/02/2016, Revision
0, Docket: PROJ0769.
12. RP-0316-17619-NP, Human Factors Engineering Human-System
Interface Design Results Summary Report, April 2019, Revision 2,
Docket: 52-048.
13. RP-0516-49116-NP, Control Room Staffing Plan Validation
Results, 12/02/2016, Revision 1, Docket: PROJ0769.
14. RP-0914-8534-NP, Human Factors Engineering Program
Management Plan, April 2019, Revision 5, Docket: 52-048.
15. RP-0914-8543-NP, Human Factors Verification and Validation
Implementation Plan, April 2019, Revision 5, Docket: 52-048.
16. RP-0914-8544-NP, Human Factors Engineering Design
Implementation Implementation Plan, November 2019, Revision 4,
Docket: 52-048, NuScale Nonproprietary.
17. RP-1018-61289-NP, Human Factors Engineering Verification and
Validation Results Summary Report, July 2019, Revision 1, Docket:
52-048.
18. RP-1215-20253-NP, Control Room Staffing Plan Validation
Methodology, 12/02/2016, Revision 3, Docket: PROJ0769.
19. TR-0116-20781-NP, Fluence Calculation Methodology and
Results, July 2019, Revision 1, Docket: 52-048.
20. TR-0116-20825-NP-A, Applicability of AREVA Fuel Methodology
for the NuScale Design, June 2016, Revision 1, Docket: PROJ0769.
21. TR-0116-21012-NP-A, NuScale Power Critical Heat Flux
Correlations, December 2018, Revision 1, Docket: PROJ0769.
22. TR-0316-22048-NP, Nuclear Steam Supply System Advanced
Sensor Technical Report, May 2020, Revision 3, Docket: 52-048.
23. TR-0515-13952-NP-A, Risk Significance Determination, October
2016, Revision 0, Docket: PROJ0769, NuScale Nonproprietary.
24. TR-0516-49084-NP, Containment Response Analysis Methodology
Technical Report, May 2020, Revision 3, Docket: 52-048.
25. TR-0516-49416-NP-A, Non-Loss-of-Coolant Accident Analysis
Methodology, July 2020, Revision 3, Docket: PROJ0769.
[[Page 35019]]
26. TR-0516-49417-NP-A, Evaluation Methodology for Stability
Analysis of the NuScale Power Module, March 2020, Revision 1,
Docket: PROJ0769.
27. TR-0516-49422-NP-A, Loss-of-Coolant Accident Evaluation
Model, July 2020, Revision 2, Docket: PROJ0769.
28. TR-0616-48793-NP-A, Nuclear Analysis Codes and Methods
Qualification, November 2018, Revision 1, Docket: PROJ0769.
29. TR-0616-49121-NP, NuScale Instrument Setpoint Methodology
Technical Report, May 2020, Revision 3, Docket: 52-048.
30. TR-0716-50350-NP-A, Rod Ejection Accident Methodology, June
2020, Revision 1, Docket: PROJ0769.
31. 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, Docket: PROJ0769.
32. TR-0716-50424-NP, Combustible Gas Control, March 2019,
Revision 1, Docket: PROJ0769.
33. TR-0716-50439-NP, NuScale Comprehensive Vibration Assessment
Program Analysis Technical Report, July 2019, Revision 2, Docket:
52-048.
34. TR-0815-16497-NP-A, Safety Classification of Passive Nuclear
Power Plant Electrical Systems, January 2018, Revision 1, Docket:
PROJ0769.
35. TR-0816-49833-NP, Fuel Storage Rack Analysis, November 2018,
Revision 1, Docket: 52-048.
36. TR-0816-50796-NP, Loss of Large Areas Due to Explosions and
Fires Assessment, June 2019, Revision 1, Docket: 52-048.
37. TR-0816-50797, Mitigation Strategies for Loss of All AC
Power Event, October 2019, Revision 3, Docket: 52-048, NuScale
Nonproprietary.
38. TR-0816-51127-NP, NuFuel-HTP2TM Fuel and Control
Rod Assembly Designs, December 2019, Revision 3, Docket: 52-048.
39. TR-0818-61384-NP, Pipe Rupture Hazards Analysis, July 2019,
Revision 2, Docket No.: 52-048.
40. TR-0915-17564-NP-A, Subchannel Analysis Methodology,
February 2019, Revision 2, Docket: PROJ0769.
41. TR-0915-17565-NP-A, Accident Source Term Methodology,
February 2020, Revision 4, Docket: PROJ0769.
42. TR-0916-51299-NP, Long-Term Cooling Methodology, May 2020,
Revision 3, Docket: 52-048.
43. TR-0916-51502-NP, NuScale Power Module Seismic Analysis,
April 2019, Revision 2, Docket: 52-048.
44. TR-0917-56119-NP, CNV Ultimate Pressure Integrity, June
2019, Revision 1, Docket No. 52-048.
45. TR-0918-60894-NP, NuScale Comprehensive Vibration Assessment
Program Measurement and Inspection Plan Technical Report, August
2019, Revision 1, Docket No.: 52-048.
46. NP-TR-1010-859-NP-A, NuScale Topical Report: Quality
Assurance Program Description for the NuScale Power Plant, May 2020,
Revision 5, Docket: PROJ0769, NuScale Nonproprietary.
47. TR-1015-18177-NP, Pressure and Temperature Limits
Methodology, October 2018, Revision 2, Docket: 52-048.
48. TR-1015-18653-NP-A, Design of the Highly Integrated
Protection System Platform, May 2017, Revision 2, Docket: PROJ0769.
49. TR-1016-51669-NP, NuScale Power Module Short-Term Transient
Analysis, July 2019, Revision 1, Docket: 52-048.
50. TR-1116-51962-NP, NuScale Containment Leakage Integrity
Assurance, May 2019, Revision 1, Docket: 52-048.
51. TR-1116-52065-NP, Effluent Release (GALE Replacement)
Methodology and Results, November 2018, Revision 1, Docket: 52-048.
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 entirely 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 criteria of 10 CFR part 20
and the requirements of 10 CFR part 50, appendix A, General Design
Criterion (GDC) 4 and GDC 31 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.
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 [DATE 120 DAYS AFTER DATE OF PUBLICATION
OF FINAL RULE IN THE Federal Register], that are applicable and
technically relevant, as described in the final safety evaluation
report.
[[Page 35020]]
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 (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 in the following table:
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 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.
9. Paragraph (c)(1) of 10 CFR 50.62--Diverse equipment to
initiate a turbine trip under conditions indicative of an
anticipated transient without scram.
10. 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.
11. Appendix A to 10 CFR part 50, GDC 19--Equipment outside the
control room with capability for cold shutdown of the reactor.
12. 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.
13. Appendix A to 10 CFR part 50, GDC 33--Reactor coolant makeup
for protection against small breaks in the reactor coolant pressure
boundary.
14. Appendix A to 10 CFR part 50, GDC 40--Periodic pressure and
functional testing of containment heat removal system.
15. Appendix A to 10 CFR part 50, GDC 52--Design to allow
periodic containment leakage rate testing.
16. 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.
17. 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;
[[Page 35021]]
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, 10, 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.
ML19303C179) 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 those site parameters
specified in the NuScale environmental report.
C. The Commission does not consider operational requirements for
an applicant or 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
October 29, 2021, 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;
[[Page 35022]]
(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 stands on an asserted noncompliance with
an ITAAC acceptance criterion in the case of a Sec. 52.103
preoperational hearing, or that the change stands 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 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 which 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.
[[Page 35023]]
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: June 25, 2021.
For the Nuclear Regulatory Commission.
Annette Vietti-Cook,
Secretary of the Commission.
[FR Doc. 2021-13940 Filed 6-30-21; 8:45 am]
BILLING CODE 7590-01-P
