[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.
---------------------------------------------------------------------------

    \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