NuScale Small Modular Reactor Design Certification, 3287-3311 [2023-00729]

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• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 88, Number 12 (Thursday, January 19, 2023)]
[Rules and Regulations]
[Pages 3287-3311]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2023-00729]



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Federal Register / Vol. 88, No. 12 / Thursday, January 19, 2023 / 
Rules and Regulations

[[Page 3287]]



NUCLEAR REGULATORY COMMISSION

10 CFR Part 52

[NRC-2017-0029]
RIN 3150-AJ98


NuScale Small Modular Reactor Design Certification

AGENCY: U.S. Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its 
regulations to certify the NuScale standard design for a small modular 
reactor. Applicants or licensees intending to construct and operate a 
NuScale standard design may do so by referencing this design 
certification rule. The applicant for certification of the NuScale 
standard design is NuScale Power, LLC.

DATES: This final rule is effective on February 21, 2023. The 
incorporation by reference of certain publications listed in the rule 
is approved by the Director of the Federal Register as of February 21, 
2023.

ADDRESSES: Please refer to Docket ID NRC-2017-0029 when contacting the 
NRC about the availability of information for this action. You may 
obtain publicly available information related to this action by any of 
the following methods:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2017-0029. Address 
questions about NRC dockets to Dawn Forder; telephone: 301-415-3407; 
email: [email protected]. For technical questions, contact the 
individuals listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin Web-based ADAMS 
Search.'' For problems with ADAMS, please contact the NRC's Public 
Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or 
by email to [email protected]. For the convenience of the reader, 
instructions about obtaining materials referenced in this document are 
provided in the ``Availability of Documents'' section.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room P1 B35, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852. To make an appointment to 
visit the PDR, please send an email to [email protected] or call 1-
800-397-4209 or 301-415-4737, between 8:00 a.m. and 4:00 p.m. (ET), 
Monday through Friday, except Federal holidays.
     Technical Library: The Technical Library, which is located 
at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 
20852, is open by appointment only. Interested parties may make 
appointments to examine documents by contacting the NRC Technical 
Library by email at [email protected] between 8:00 a.m. and 4:00 
p.m. (ET), Monday through Friday, except Federal holidays.

FOR FURTHER INFORMATION CONTACT: Yanely Malave, Office of Nuclear 
Material Safety and Safeguards, telephone: 301-415-1519, email: 
[email protected], and Carolyn Lauron, Office of Nuclear Reactor 
Regulation, telephone: 301-415-2736, email: [email protected]. 
Both are staff of the U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.

SUPPLEMENTARY INFORMATION:

Table of Contents

I. Background
II. Opportunities for Public Participation
III. Regulatory and Policy Issues
IV. Technical Issues Associated With the NuScale Design
V. Discussion
    A. Introduction (Section I)
    B. Definitions (Section II)
    C. Scope and Contents (Section III)
    D. Additional Requirements and Restrictions (Section IV)
    E. Applicable Regulations (Section V)
    F. Issue Resolution (Section VI)
    G. Duration of This Appendix (Section VII)
    H. Processes for Changes and Departures (Section VIII)
    I. [Reserved] (Section IX)
    J. Records and Reporting (Section X)
VI. Public Comment Analysis
VII. Section-by-Section Analysis
VIII. Regulatory Flexibility Certification
IX. Regulatory Analysis
X. Backfitting and Issue Finality
XI. Plain Writing
XII. Environmental Assessment and Finding of No Significant Impact
XIII. Paperwork Reduction Act
XIV. Congressional Review Act
XV. Agreement State Compatibility
XVI. Voluntary Consensus Standards
XVII. Availability of Documents
XVIII. Incorporation by Reference--Reasonable Availability to 
Interested Parties

I. Background

    Part 52 of title 10 of the Code of Federal Regulations (10 CFR), 
``Licenses, Certifications, and Approvals for Nuclear Power Plants,'' 
subpart B, ``Standard Design Certifications,'' presents the process for 
obtaining standard design certifications. By letter dated December 31, 
2016, NuScale Power, LLC, (NuScale Power) filed its application for 
certification of the NuScale standard design (hereafter referred to as 
NuScale). The NRC published a notification of receipt of the design 
certification application (DCA) in the Federal Register on February 22, 
2017 (82 FR 11372). On March 30, 2017, the NRC published a notification 
of acceptance for docketing of the application in the Federal Register 
(82 FR 15717) and assigned docket number 52-048. The preapplication 
information submitted before the NRC formally accepted the application 
can be found in ADAMS under Docket No. PROJ0769.
    NuScale is the first small modular reactor design reviewed by the 
NRC. NuScale is based on a small light water reactor developed at 
Oregon State University in the early 2000s. It consists of one or more 
NuScale power modules (hereafter referred to as power module(s)). A 
power module is a natural circulation light water reactor composed of a 
reactor core, a pressurizer, and two helical coil steam generators 
located in a common reactor pressure vessel that is housed in a compact 
cylindrical steel containment. The NuScale reactor building is designed 
to hold up to 12 power modules. Each power module has a rated thermal 
output of 160 megawatt thermal (MWt) and electrical output of 50 
megawatt electric (MWe), yielding a

[[Page 3288]]

total capacity of 600 MWe for 12 power modules. All the NuScale power 
modules are partially submerged in a common safety-related pool, which 
is also the ultimate heat sink for up to 12 power modules. The pool 
portion of the reactor building is located below grade. The design 
utilizes several first-of-a-kind approaches for accomplishing key 
safety functions, resulting in no need for Class 1E safety-related 
power (no emergency diesel generators), no need for pumps to inject 
water into the core for post-accident coolant injection, and reduced 
need for control room staffing while providing safe operation of the 
plant during normal and post-accident operation.

II. Opportunities for Public Participation

    The proposed rule and environmental assessment were published in 
the Federal Register on July 1, 2021, for a 60-day public comment 
period (86 FR 34999). The public comment period was scheduled to close 
on August 30, 2021. The NRC subsequently extended the comment period by 
45 days (86 FR 47251; August 24, 2021), providing a total comment 
period of 105 days. The public comment period closed on October 14, 
2021. The public comments informed the development of this final rule.

III. Regulatory and Policy Issues

A. Exemptions for Future Applicants Referencing NuScale

1. Control Room Staffing Requirements
    The requirements in Sec. Sec.  50.54(k) and 50.54(m) identify the 
minimum number of licensed operators that must be on site, in the 
control room, and at the controls. The requirements are conditions in 
every nuclear power reactor operating license issued under 10 CFR part 
50, ``Domestic Licensing of Production and Utilization Facilities.'' 
The requirements also are conditions in every combined license (COL) 
issued under 10 CFR part 52; however, they are applicable only after 
the Commission makes the finding under Sec.  52.103(g) that the 
acceptance criteria in the COL are met.
    In a letter to the NRC, dated September 15, 2015, NuScale Power 
proposed that 6 licensed operators would operate up to 12 power modules 
from a single control room. The staffing proposal would meet the 
requirements of Sec.  50.54(k) but would not meet the requirements in 
Sec.  50.54(m)(2)(i) because the minimum requirements for the onsite 
staffing table in Sec.  50.54(m)(2)(i) do not address operation of more 
than two units from a single control room. The proposal also would not 
meet Sec.  50.54(m)(2)(iii), which requires a licensed operator at the 
controls for each fueled unit. Absent alternative staffing 
requirements, future applicants referencing the NuScale design would 
need to request an exemption.
    In DCA, Part 7, Section 6, NuScale requested that the NRC approve 
design-specific control room staffing requirements in lieu of the 
requirements in Sec.  50.54(m). In the DCA Part 7, Section 6.2, 
``Justification for Rulemaking,'' NuScale Power provided a technical 
basis for its proposed alternative control room staffing requirements. 
NuScale Power's proposed approach is consistent with SECY-11-0098, 
``Operator Staffing for Small or Multi-Module Nuclear Power Plant 
Facilities,'' dated July 22, 2011. For the reasons described in Chapter 
18, Section 18.5.4.2, ``Evaluation of the Applicant's Technical 
Basis,'' of the final safety evaluation report, the NRC found that 
NuScale Power's proposed staffing level, as described in the DCA Part 
7, Section 6, is acceptable. Because Section V, ``Applicable 
Regulations,'' of this final rule includes the alternative staffing 
requirement provisions, staffing table, and appropriate table notes, a 
future applicant or licensee that references appendix G to 10 CFR part 
52 will not need to request an exemption from Sec.  50.54(m).
2. Preoperational and Periodic Testing of Primary Reactor Containment
    General Design Criterion (GDC) 52, ``Capability for Containment 
Leakage Rate Testing,'' requires that the containment be designed so 
that periodic, integrated leakage rate testing can be conducted at 
containment design pressure; the underlying purpose of which is to 
provide design capability for testing that assures that containment 
leakage integrity is maintained and containment vessel leakage does not 
exceed allowable leakage rate values (see appendix J to 10 CFR part 
50). Under 10 CFR 50.54(o), operating licenses and combined licenses 
for certain water-cooled power reactors must include a condition that 
the primary containment shall be subject to appendix J to 10 CFR part 
50, ``Primary Reactor Containment Leakage Testing for Water-Cooled 
Power Reactors.'' Appendix J to 10 CFR part 50 requires that primary 
reactor containments meet the containment leakage test requirements to 
provide for preoperational and periodic verification by tests of the 
leak-tight integrity of the primary reactor containment (Type A) and 
systems and components that penetrate containment (Type B and Type C).
    NuScale Power requested an exemption from GDC 52 in order to not 
design NuScale to include the capability for Type A testing and 
requested that the design certification rule exempt licensees 
referencing the NuScale design certification rule from the requirement 
for Type A testing in appendix J to 10 CFR part 50. NuScale Power's 
request was based on the NuScale small modular reactor design meeting 
the underlying purpose of the regulation through means not anticipated 
when the NRC issued GDC 52 and appendix J to 10 CFR part 50. NuScale 
Power stated that the NuScale containment has two primary features 
distinguishing it from containments at existing light water reactors 
that provide assurance that no unknown leakage pathways will be 
present. First, the NuScale containment is designed and would be 
constructed as a pressure vessel, and therefore leakage due to vessel 
design or fabrication flaws would be identified during a required 
preservice structural integrity test. In contrast to a Type A test, 
this test is a hydrostatic leakage test at design pressure, with no 
visible leakage as its acceptance criterion. Second, the containment is 
100-percent inspectable, both inside and outside, whereby aging-related 
flaws leading to potential leakage could be observed. Containment 
leakage integrity assurance for NuScale is described in detail in 
technical report TR-1116-51962-NP, ``NuScale Containment Leakage 
Integrity Assurance,'' Rev. 1 (May 2019), which this final rule 
incorporates by reference. NuScale Power stated that the required 
preservice tests and inservice inspections described in TR-1116-51962-
NP, including Type B and Type C testing without Type A testing, ensure 
that containment leakage rates remain acceptable.
    In Chapter 6, Section 6.2.6.4, ``Technical Evaluation for Exemption 
Request No. 7,'' of the final safety evaluation report, the NRC staff 
concluded that granting this exemption from Type A testing, and 
associated design features required by GDC 52 to provide for Type A 
testing, is acceptable because the NuScale design relies on the 
preservice pressure test, successful Type B and C testing at each 
refueling as required in appendix J to 10 CFR part 50, periodic 
inservice inspections, and direct observation of the entire vessel to 
identify potential degradation or unknown leakage pathways for the 
remainder of the service life for the containment.

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    The NRC received a comment that the exemption from the requirement 
for Type A testing in appendix J to 10 CFR part 50 should have been 
listed in the proposed rule. The NRC agrees that the exemption should 
have been included in the proposed rule. The NRC's conclusion that Type 
A testing is not necessary for NuScale was noticed for comment as the 
basis for the exemption from GDC 52. The exemption from Type A testing 
itself was discussed in detail in the same section of final safety 
evaluation report that evaluated the exemption from GDC 52. Although 
the exemption from Type A testing was not included in the proposed 
rule, the change to this final rule only specifies that future 
licensees that reference this final rule will not be required to 
perform Type A testing for which NuScale is not designed or required to 
be capable of. Therefore, the NRC concludes that the exemption from the 
Type A test in appendix J to 10 CFR part 50 is a logical outgrowth of 
the proposed rule. In addition, because the issue of whether Type A 
testing is necessary for NuScale was noticed in the proposed rule and 
the NRC received no comments on the matter, the NRC finds that notice 
and comment on this exemption from Type A testing is unnecessary within 
the meaning of 5 U.S.C. 553(b).
    Thus, Section V, ``Applicable Regulations,'' in this final rule 
includes an exemption for licensees referencing appendix G to 10 CFR 
part 52 from the requirement of appendix J to 10 CFR part 50 to conduct 
Type A testing.

B. Incorporation by Reference

    Section III.A, ``Incorporation by reference approval,'' of appendix 
G to 10 CFR part 52 lists documents that were approved by the Director 
of the Office of the Federal Register for incorporation by reference 
into this appendix. Section III.B.2 identifies information that is not 
within the scope of the design certification and, therefore, is not 
incorporated by reference into this appendix. This information includes 
conceptual design information, as defined in Sec.  52.47(a)(24), and 
the discussion of ``first principles'' described in the Design Control 
Document (DCD) Part 2, Tier 2, Section 14.3.2, ``Tier 1 Design 
Description and Inspections, Tests, Analyses, and Acceptance Criteria 
First Principles.''
    The final rule has been updated to align with the Office of the 
Federal Register's latest guidance for incorporation by reference, 
issued on March 1, 2022, as supplemented by Release 1-2022 to the 
Incorporation by Reference Handbook.

C. Issues Not Resolved by the Design Certification

    The NRC identified three issues as not resolved within the meaning 
of Sec.  52.63(a)(5). There was insufficient information available for 
the NRC to resolve issues regarding (1) the shielding wall design in 
certain areas of the plant, (2) the potential for containment leakage 
from the combustible gas monitoring system, and (3) the ability of the 
steam generator tubes to maintain structural and leakage integrity 
during density wave oscillations in the secondary fluid system, 
including the method of analysis to predict the thermal-hydraulic 
conditions of the steam generator secondary fluid system and resulting 
loads, stresses, and deformations from density wave oscillations from 
reverse flow.
1. Shielding Wall Design
    As discussed in Section 12.3.4.1.2 of the final safety evaluation 
report, the NRC found that there were insufficient design details 
available regarding shielding wall design with the presence of large 
penetrations, such as the main steam lines; main feedwater lines; and 
power module bay heating, ventilation, and air conditioning lines in 
the radiation shield wall between the power module bay and the reactor 
building steam gallery area. Without this shielding design information, 
the NRC is unable to confirm that the radiological doses to workers 
will be maintained within the radiation zone limits specified in the 
application.
    This issue is narrowly focused on the shielding walls between the 
reactor module bays and the reactor building steam gallery areas. The 
radiation zones and dose calculations, including dose calculations for 
the dose to workers, members of the public, and environmental 
qualification, in areas outside of the reactor module bay are 
calculated assuming a solid wall and currently do not account for 
penetrations in the shield wall. An applicant is required to 
demonstrate penetration shielding adequate to address the following 
issues in the NuScale DCD: the plant radiation zones, environmental 
qualification dose calculations, and dose estimates for workers and the 
public. An applicant can provide this information for the NRC to review 
because this issue involves a localized area of the plant without 
affecting other aspects of the NRC's review of the NuScale design. 
Therefore, the NRC has determined that this information can be provided 
by an applicant that references this appendix without a demonstrable 
impact on safety or standardization. Appendix G to 10 CFR part 52, 
Section VI, ``Issue Resolution,'' clarifies that this issue is not 
resolved within the meaning of Sec.  52.63(a)(5), and Section IV, 
``Additional Requirements and Restrictions,'' states that the COL 
applicant is responsible for providing the design information to 
address this issue.
2. Containment Leakage From the Combustible Gas Monitoring System
    As documented in Section 12.3.4.1.3 of the final safety evaluation 
report, there was insufficient information available regarding the 
NuScale combustible gas monitoring system and the potential for leakage 
from this system outside containment. Without additional information 
regarding the potential for leakage from this system, the NRC was 
unable to determine whether this leakage could impact analyses 
performed to assess main control room dose consequences, offsite dose 
consequences to members of the public, and whether this system can be 
safely re-isolated after monitoring is initiated due to potentially 
high dose levels at or near the isolation valve location. The isolation 
valve can only be operated locally, and dose levels at the valve 
location have not been determined.
    This issue is narrowly focused on the radiation dose implications 
as a result of using the post-accident combustible gas monitoring loop. 
An applicant is required under Sec. Sec.  50.34(f)(2) and 52.47(a)(2) 
to demonstrate either that offsite and main control room dose 
calculations are not exceeded or that the system can be safely re-
isolated, if needed. This issue does not affect normal plant operation 
or non-core damage accidents. The issue may be resolved by performing 
radiation dose calculations and demonstrating that doses would remain 
within applicable dose limits in 10 CFR part 20, ``Standards for 
Protection Against Radiation.'' More information may be available at 
the application stage that would allow for more detailed calculations. 
Any design changes to address this issue would only affect the 
combustible gas monitoring loop to ensure it can be re-isolated or to 
ensure that dose limits are not exceeded. Such design changes likely 
would not have an impact on other systems or equipment, and the NRC 
would review such changes and any resulting effects on other 
structures, systems, and components during the application review to 
determine whether there is reasonable assurance of adequate

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protection of public health and safety. Therefore, the NRC has 
determined that this information can be provided by an applicant that 
references this appendix without a demonstrable impact on safety or 
standardization. Appendix G to 10 CFR part 52, Section VI, ``Issue 
Resolution,'' clarifies that this issue is not resolved within the 
meaning of Sec.  52.63(a)(5), and Section IV, ``Additional Requirements 
and Restrictions,'' states that the COL applicant is responsible for 
providing the design information to address this issue.
3. Steam Generator Stability During Density Wave Oscillations and 
Associated Method of Analysis
    Section 5.4.1.2, ``System Design,'' in Revision 2 of the DCA Part 
2, Tier 2 (ADAMS Accession No. ML18310A345), stated that a flow 
restriction device at the inlet to each steam generator tube ``ensures 
secondary-side flow stability and precludes density wave 
oscillations.'' However, the applicant modified this section in 
Revision 3 of the DCA Part 2, Tier 2 (ADAMS Accession No. ML19241A431), 
to state that the steam generator inlet flow restrictors provide the 
necessary secondary-side pressure drop ``to reduce flow oscillations to 
acceptable limits.'' Revision 4.1 of the DCA (ADAMS Accession No. 
ML20205L562) revised Section 5.4.1.2 to state that the steam generator 
inlet flow restrictors are designed ``to reduce the potential for 
density wave oscillations.'' Revision 5 of this section of the DCA 
(ADAMS Accession No. ML20225A071) provides only editorial changes to 
Revision 4.1 and does not change the technical content or conclusions.
    Sections 3.9.2, 3.9.5, and 5.4.1 of the final safety evaluation 
report relied on the applicant's statements in Revision 2 and Revision 
3 of the DCA that flow oscillations in the secondary fluid system of 
the steam generators would either be precluded or minimal. After 
issuance of the advanced safety evaluation report, the NRC noted 
inconsistencies and gaps in the information provided in Sections 3.9.1, 
3.9.2, and 5.4.1 of Revision 4.1 of the DCA Part 2, Tier 2, regarding 
the potential for significant density wave oscillations in the steam 
generator tubes, including both forward and reverse secondary flow. The 
testing performed by the applicant on various conceptual designs of the 
steam generator inlet flow restrictors only involved flow in the 
forward direction without oscillation or reverse flow.
    As a result, NuScale Power has not demonstrated that the flow 
oscillations that are predicted to occur on the secondary side of the 
steam generators will not cause failure of the inlet flow restrictors. 
Structural and leakage integrity of the inlet flow restrictors in the 
steam generators is necessary to avoid damage to multiple steam 
generator tubes, caused directly by broken parts or indirectly by 
unexpected density wave oscillation loads. Damage to multiple steam 
generator tubes could disrupt natural circulation in the reactor 
coolant pathway and interfere with the decay heat removal system and 
the emergency core cooling system, which is relied upon to cool the 
reactor core in a NuScale power module. The failure of multiple steam 
generator tubes resulting from failure of an inlet flow restrictor has 
not been included within the scope of the NuScale accident analyses in 
DCA Part 2, Tier 2, Chapter 15. Therefore, the NRC concludes that 
NuScale Power has not demonstrated compliance with 10 CFR 
52.47(a)(2)(iv) and appendix A to 10 CFR part 50, GDC 4 and GDC 31, 
relative to potential impacts on steam generator tube integrity from 
inlet flow restrictor failure.
    As described previously, NuScale Power made a change to the 
description of inlet flow restrictor performance beginning with DCA 
Part 2, Tier 2, Revision 3, that indicates that the design no longer 
precludes density wave oscillations in the secondary side of the steam 
generators. As a result, the design needs a method of analysis to 
predict the thermal-hydraulic conditions of the steam generator 
secondary fluid system and resulting loads, stresses, and deformations 
from density wave oscillations including reverse flow. However, as 
described in the next paragraph, NuScale power did not provide 
verification and validation for its proposed method of analysis to 
demonstrate it is appropriate for this purpose.
    The DCA Part 2, Tier 2, Section 3.9.1.2, ``Computer Programs Used 
in Analyses,'' lists the computer programs used by NuScale Power in the 
dynamic and static analyses of mechanical loads, stresses, and 
deformations, and in the hydraulic transient load analyses of seismic 
Category I components and supports for the NuScale nuclear power plant. 
Section 3.9.1.2 states that NRELAP5 is NuScale's proprietary system 
thermal-hydraulics code for use in safety-related design and analysis 
calculations and is pre-verified and configuration-managed. The 
advanced safety evaluation report, Section 3.9.1.4.9, ``Computer 
Programs Used in Analyses,'' states that the NRELAP5 computer program 
had received verification and validation. Following preparation of the 
advanced safety evaluation report, the NRC noted a discrepancy between 
two statements in the DCA about validation for NRELAP5: DCA Part 2, 
Tier 2, Section 5.4.1.3, in Revision 4 stated that NRELAP5 was 
validated for determining density wave oscillation thermal-hydraulic 
conditions, referring to Section 15.0.2 for more information, but 
neither Section 15.0.2 nor technical report TR-1016-51669-NP describe 
validation for determining density wave oscillation thermal-hydraulic 
conditions.
    On June 19, 2020, NuScale submitted Revision 4.1 of the DCA Part 2, 
Tier 2 (ADAMS Accession No. ML20205L562; subsequently included in 
Revision 5 of the DCA submitted on July 29, 2020 (ADAMS Accession No. 
ML20225A071)), to correct the discrepancies and acknowledge the need 
for a COL applicant to address secondary-side instabilities in the 
steam generator design. Specifically, the update to Section 3.9.1.2 in 
Revision 4.1 of DCA Part 2, Tier 2, references DCA Part 2, Tier 2, 
Section 15.0.2, ``Review of Transient and Accident Analysis Methods,'' 
for the discussion of the development, use, verification, validation, 
and code limitations of the NRELAP5 computer program for application to 
transient and accident analyses. The correction to Section 3.9.1.2 also 
references technical report TR-1016-51669-NP, ``NuScale Power Module 
Short-Term Transient Analysis,'' incorporated by reference in DCA Part 
2, Tier 2, Table 1.6-2, for application of the NRELAP5 computer program 
to short-term transient dynamic mechanical loads, such as pipe breaks 
and valve actuations. In addition, the correction to Section 3.9.1.2 
includes a new COL item specifying that a COL applicant that references 
the NuScale DCD will develop an evaluation methodology for the analysis 
of secondary-side instabilities in the steam generator design. The COL 
item states that this methodology would address the identification of 
potential density wave oscillations in the steam generator tubes and 
qualification of the applicable portions of the reactor coolant system 
integral reactor pressure vessel and steam generator given the 
occurrence of density wave oscillations, including the effects of 
reverse fluid flows within the tubes. These corrections to the DCA 
clarify that the evaluation methodology for the analysis of secondary-
side instabilities in the steam generator design was not verified and 
validated as

[[Page 3291]]

part of the NuScale DCA but will need to be established by the COL 
applicant.
    This steam generator design issue is narrowly focused on the 
effects of density wave oscillations in the secondary fluid system on 
steam generator tubes to maintain structural and leakage integrity, 
including the method of analysis to predict the thermal-hydraulic 
conditions of the steam generator secondary fluid system and resulting 
loads, stresses, and deformations from density wave oscillations 
including reverse flow. No other reactor safety aspect of the steam 
generators is impacted by this design issue. As a result, the NRC finds 
that this is an isolated issue that does not affect other aspects of 
the NRC's review of the design of the NuScale nuclear power plant. 
Therefore, the NRC has determined that this information can be provided 
by an applicant that references this appendix, consistent with the 
other design information regarding steam generator integrity described 
in DCA Part 2, Tier 2, Sections 3.9.1, 3.9.2, and 5.4.1, without a 
demonstrable impact on safety or standardization. Therefore, appendix G 
to 10 CFR part 52, Section VI, ``Issue Resolution,'' clarifies that 
this issue is not resolved within the meaning of Sec.  52.63(a)(5), and 
Section IV, ``Additional Requirements and Restrictions,'' states that 
the COL applicant is responsible for providing the design information 
to address this issue.

D. The Term ``Multi-Unit'' as Applied to NuScale

    In a letter response to NuScale Power dated October 25, 2016, the 
NRC staff explained how the staff's review of NuScale would apply the 
definitions for ``nuclear power unit'' from Appendix A to 10 CFR part 
50, ``General Design Criteria for Nuclear Power Plants,'' and ``modular 
design'' from Sec.  52.1, ``Definitions.'' As defined in Appendix A to 
10 CFR part 50, a nuclear power unit is the combination of a nuclear 
reactor and the equipment for power generation. As defined in Sec.  
52.1, modular design means that the nuclear power station consists of 
two or more essentially identical nuclear reactors (modules) and that 
each module is capable of operation independent of the other modules, 
even if they have some shared systems.
    The NuScale modular design combines one or more nuclear reactors 
(up to 12) with the necessary equipment for power generation, such that 
each separate nuclear reactor can be operated independent of the stage 
of completion or operating condition of any other nuclear reactor on 
the same site. Therefore, each reactor (i.e., power module) is a 
separate nuclear power unit. However, NuScale's modular design means 
that some multi-unit considerations are integral to the design. The 
NuScale DCD addresses multi-unit considerations other than construction 
for up to 12 power modules in a single reactor building, but the 
NuScale DCD does not address multi-unit issues that may arise if a 
NuScale facility is constructed and operated on the same site as 
another nuclear facility.
    For previously certified or licensed power reactor designs (one 
nuclear power unit per reactor building), multi-unit site 
considerations arose when multiple nuclear power units (in separate 
reactor buildings) on the same site could affect the construction or 
operation of another unit in a manner not previously reviewed by the 
NRC. However, because the NuScale design has been reviewed and is 
certified for multiple units in a single reactor building, issues 
related to multiple NuScale units in the same reactor building 
constructed at the same time have been resolved. Future applicants 
referencing the NuScale design certification will need to address 
multi-unit construction issues and, if applicable, multi-unit issues 
for a proposed NuScale facility to be constructed and operated on the 
same site as another nuclear facility, including adding additional 
NuScale modules to a previously licensed NuScale reactor building.
    The NRC has added a definition of the term ``nuclear power unit'' 
to this final rule.

IV. Technical Issues Associated With the NuScale Design

    The NRC identified significant technical issues associated with the 
following design areas that were resolved during the review:
     Comprehensive vibration assessment program;
     Containment safety analysis;
     Emergency core cooling system inadvertent actuation block 
valve;
     Conformance with GDC 27, ``Combined Reactivity Control 
Systems Capability,'' of appendix A, ``General Design Criteria for 
Nuclear Power Plants,'' to 10 CFR part 50;
     Absence of safety-related Class 1E alternating current 
(AC) or direct current (DC) electrical power;
     Accident source term methodology;
     Boron redistribution during passive cooling modes.
    In addition, the NRC granted 17 exemptions from 10 CFR part 50 to 
address various aspects of NuScale Power's design.

A. Comprehensive Vibration Assessment Program

    The NuScale comprehensive vibration assessment program limits 
potentially adverse effects from flow, acoustic, and mechanically 
induced vibrations and resonances on NuScale power module components, 
including the helical coil steam generators. The NuScale steam 
generators are different from those of operating pressurized-water 
reactors in that the primary reactor coolant is on the outside of the 
steam generator tubes and the steam is on the inside. Because of this 
design, there is the possibility of density wave oscillation 
instabilities in the secondary coolant, which could challenge the 
integrity of the tubes. The NRC's review and findings, including 
independent analyses and observation of vibration testing, are 
documented in detail in Chapter 3, ``Design of Structures, Systems, 
Components and Equipment,'' Section 3.9.2, ``Dynamic Testing and 
Analysis of Systems, Structures, and Components,'' of the final safety 
evaluation report. The review focused on assuring that the design of 
the helical coil steam generator tubes would not result in issues with 
flow-induced vibration.
    As part of the comprehensive vibration assessment, the NRC also 
reviewed and found acceptable the steam generator tube margin against 
fluid-elastic instability, steam generator tube margin against vortex 
shedding, control rod drive shaft margin against vortex shedding, in-
core instrument guide tube against vortex shedding, decay heat removal 
system piping against acoustic resonance, and control rod assembly 
guide tube against turbulence buffeting. The steam generator tube 
margins against fluid-elastic instability and vortex shedding will be 
validated in the TF-3 testing facility as described in DCA Part 2, Tier 
1, Section 2.1.1, ``Design Description.'' In addition, the initial 
startup testing will confirm that flow-induced vibration will not cause 
adverse effects on the plant system components including the steam 
generator tubes. With the exception of the steam generator tube and 
inlet flow restrictor issue discussed in Section III.C.3, the NRC found 
the comprehensive vibration assessment program adequate to ensure the 
structural integrity of the NuScale power module components.

B. Containment Safety Analysis

    NuScale incorporates novel and unique features that result in 
transient thermal-hydraulic responses that are different from those of 
currently licensed reactors.

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    There are several peak containment pressure analysis technical 
issues unique to NuScale, including the associated thermal-hydraulic 
analyses. In support of containment safety analysis, NuScale Power 
submitted technical report TR-0516-49084-NP, Revision 3, ``Containment 
Response Analysis Methodology,'' May 2020, which describes the 
conservative containment pressure and temperature safety analyses for 
several design-basis events related to the containment design margins. 
NuScale Power also submitted topical report TR-0516-49422-NP, ``Loss-
of-Coolant Accident Evaluation Model,'' Revision 1, dated November 
2019. This topical report describes the evaluation model used to 
analyze the power module response during a design-basis loss-of-coolant 
accident. The NRC reviewed this topical report as part of the 
containment safety analysis.
    The NRC also observed thermal-hydraulic performance testing at 
NuScale Power's integrated system test facility, which validates the 
analytical model. Based on initial testing results and thermal-
hydraulic analyses, NuScale Power made design changes to increase the 
initial reactor building pool level and the in-containment vessel 
design pressure to account for some uncertainties.
    The NRC reviewed the details of the computer thermal-hydraulic 
evaluation model described in the DCA Part 2, Tier 2, Section 6.2.1.1, 
to determine whether any uncertainties were properly accounted for and 
found the containment design margins to be acceptable. The associated 
safety evaluation report approving topical report TR-0516-49422 was 
issued on February 18, 2020. The NRC's review and specific findings, 
including independent analyses and observation of NuScale testing, are 
documented in Chapter 6, ``Engineered Safety Features,'' Section 
6.2.1.1, ``Containment Structure,'' of the safety evaluation report.

C. Emergency Core Cooling System Inadvertent Actuation Block Valve

    The NuScale emergency core cooling system relies on natural 
circulation cooling of the reactor core by releasing the heated reactor 
coolant steam from the top of the reactor pressure vessel through three 
reactor vent valves into the containment vessel and returning the 
cooled condensed reactor coolant water to the reactor pressure vessel 
through two reactor recirculation valves. Each reactor vent valve and 
reactor recirculation valve consists of a first-of-a-kind arrangement 
of a main valve, an inadvertent actuation block (IAB) valve, a solenoid 
trip valve, and a solenoid reset valve. The IAB valve for each reactor 
vent valve and reactor recirculation valve is designed to close rapidly 
to prevent its corresponding emergency core cooling system main valve 
from opening when the reactor coolant system is at high pressure 
conditions. Premature opening of the emergency core cooling system main 
valves could result in fuel damage. The IAB valve then opens at reduced 
reactor coolant system pressure to allow the main valve to open and 
permit natural circulation cooling of the reactor core in response to a 
plant event. Although the valve assemblies are considered an active 
component, NuScale Power does not apply the single failure criterion to 
the IAB valve, including to the IAB valve's function to close. 
Consistent with Commission safety goals and the practice of risk-
informed decisionmaking, the NRC evaluated the NuScale emergency core 
cooling system valve system without assuming a single active failure of 
the IAB valve to close.
    During design demonstration tests of the first-of-a-kind emergency 
core cooling system valve system performed under Sec.  50.43(e), 
NuScale Power implemented design modifications to the main valve and 
IAB valve to demonstrate that the IAB valve will operate within a 
specific design pressure range. The DCD specifies that the emergency 
core cooling system valves (including the IAB valves) will be qualified 
under American Society of Mechanical Engineers Standard QME-1-2007, 
``Qualification of Active Mechanical Equipment Used in Nuclear Power 
Plants,'' as endorsed by NRC Regulatory Guide 1.100, Revision 3, 
``Seismic Qualification of Electrical and Active Mechanical Equipment 
and Functional Qualification of Active Mechanical Equipment for Nuclear 
Power Plants,'' prior to installation in a NuScale nuclear power plant. 
Additionally, the NRC regulations in Sec.  50.55a require that a 
NuScale nuclear power plant meet the requirements of the American 
Society of Mechanical Engineers Operation and Maintenance of Nuclear 
Power Plants, Division 1, OM Code: Section IST (OM Code) as 
incorporated by reference in Sec.  50.55a for inservice testing of the 
emergency core cooling system valves, unless relief is granted or an 
alternative is authorized by the NRC. The NRC's review and findings 
related to the IAB valve are documented in safety evaluation report 
Chapter 3, ``Design of Structures, Systems, Components and Equipment,'' 
Section 3.9.6, ``Functional Design, Qualification, and Inservice 
Testing Programs for Pumps, Valves, and Dynamic Restraints.'' These 
findings show that the NRC regulatory requirements and DCD Part 2, Tier 
2 provisions provide reasonable assurance that the emergency core 
system valve system will be capable of performing its design-basis 
functions in light of the safety significance of the required opening 
and closing pressures for the individual IAB valves.
    Further, Chapter 15, ``Transient and Accident Analyses,'' Section 
15.0.0.5, ``Limiting Single Failures,'' of the safety evaluation report 
states that the IAB valve is a first-of-a-kind, safety-significant, 
active component integral to the NuScale emergency core cooling system. 
NuScale Power does not apply the single failure criterion to the IAB 
valve, and, on July 2, 2019, the Commission directed the staff in SRM-
SECY-19-0036, ``Staff Requirements--SECY-19-0036--Application of the 
Single Failure Criterion to NuScale Power LLC's Inadvertent Actuation 
Block Valves,'' to ``review Chapter 15 of the NuScale Design 
Certification Application without assuming a single active failure of 
the inadvertent actuation block valve to close.'' The Commission 
further stated that ``[t]his approach is consistent with the 
Commission's safety goal policy and associated core damage and large 
release frequency goals and existing Commission direction on the use of 
risk-informed decision-making, as articulated in the 1995 Policy 
Statement on the Use of Probabilistic Risk Assessment Methods in 
Nuclear Regulatory Activities and the White Paper on Risk-Informed and 
Performance-Based Regulation (in SRM-SECY-98-144, ``White Paper on 
Risk-Informed and Performance-Based Regulation,'' and Yellow 
Announcement 99-019).''
    Based on the NRC's historic application of the single failure 
criterion and Commission direction on the subject, as described in 
SECY-77-439, ``Single Failure Criterion''; SRM-SECY-94-084, ``Policy 
and Technical Issues associated with the Regulatory Treatment of Non-
Safety Systems and Implementation of Design Certification and Light-
Water Reactor Design Issues''; and SRM-SECY-19-0036, the NRC has 
retained discretion, in fact or application-specific circumstances, to 
decide when to apply the single failure criterion. The Commission's 
decision in SRM-SECY-19-0036 provides direction regarding the 
appropriate application and interpretation of the regulatory 
requirements in 10 CFR part 50 to the NuScale IAB valve's function to 
close. This decision is similar to those in

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previous Commission documents that addressed the use of the single 
failure criterion and provided clarification on when to apply the 
single failure criterion in other specific instances.

D. Conformance With General Design Criterion 27, ``Combined Reactivity 
Control Systems Capability''

    NuScale Power determined that, under certain end-of-cycle scenarios 
with one control rod stuck out, the NuScale reactivity control systems 
could not prevent re-criticality and return to power. This result does 
not meet GDC 27 of appendix A to 10 CFR part 50, which covers 
reactivity control systems to reliably control reactivity changes under 
postulated accident conditions with margin for stuck control rods. 
Therefore, NuScale Power submitted an exemption request for GDC 27 
(refer to Section 15, ``10 CFR 50, Appendix A, Criterion 27, `Combined 
Reactivity Control Systems Capability,' '' of DCA Part 7, 
``Exemptions'').
    NuScale Power analyses determined that the specified acceptable 
fuel design limits would not be exceeded and that core cooling would be 
maintained during a return to power under these scenarios. The global 
core power level would be less than 10 percent and within capacity of 
the safety-related, passive decay heat removal system. The NRC 
independently verified NuScale Power's results and found that NuScale 
achieves the fundamental safety functions for nuclear reactor safety, 
which are to control heat generation, remove heat, and limit the 
release of radioactive materials. Chapter 15, Section 15.0.6.4.1, of 
the safety evaluation report contains details of the evaluation of this 
exemption request. Additional information is provided in SECY-18-0099, 
``NuScale Power Exemption Request from 10 CFR part 50, Appendix A, 
General Design Criterion 27, `Combined Reactivity Control Systems 
Capability,' '' dated October 9, 2018. The NRC granted the exemption 
request.

E. Absence of Safety-Related Class 1E AC or DC Electrical Power

    NuScale does not contain safety-related Class 1E AC or DC 
electrical power systems. The purpose of appendix A to 10 CFR part 50, 
GDC 17, ``Electric Power Systems,'' is to ensure that sufficient 
electric power is available to accomplish plant functions important to 
safety. NuScale provides passive safety systems and features to 
accomplish plant safety-related functions without reliance on 
electrical power.
    NuScale incorporates several innovative features that reduce the 
overall complexity of the design and lower the number of safety-related 
systems necessary to mitigate postulated accidents. NuScale has no 
safety-related functions that rely on electrical power. For example, 
the emergency core cooling system performs its safety function without 
reliance on safety-related electrical power or external sources of 
coolant inventory makeup. NuScale Power provided a methodology to 
substantiate its assertion that the safety-related systems do not rely 
on Class 1E electrical power in topical report TR-0815-16497, Revision 
1, ``Safety Classification of Passive Nuclear Power Plant Electrical 
Systems,'' dated February 7, 2017. The NRC reviewed topical report TR-
0815-16497 and concluded that NuScale Power demonstrated that the 
safety-related systems do not rely on Class 1E electrical power. The 
NRC's review and conclusions are documented in a safety evaluation 
report approving topical report TR-0815-16497, issued December 13, 
2017, as described in the final safety evaluation report for Chapter 1, 
``Introduction and General Discussion,'' and included in the approved 
version of the topical report, TR-0815-16497-NP-A.
    Because no safety-related functions of NuScale rely on electrical 
power, NuScale does not need any safety-related electrical power 
systems. Therefore, NuScale Power requested an exemption from GDC 17, 
which requires the provision of onsite and offsite power to provide 
sufficient capacity and capability to assure that (1) specified 
acceptable fuel design limits and design conditions of the reactor 
coolant pressure boundary are not exceeded as a result of anticipated 
operational occurrences and (2) the core is cooled and containment 
integrity and other vital functions are maintained in the event of 
postulated accidents. The NRC determined that, subject to limitations 
and conditions stipulated in its safety evaluation report for TR-0815-
16497, the underlying purpose of GDC 17 (to ensure sufficient electric 
power is available to accomplish the safety functions of the respective 
systems), is met without reliance on Class 1E electric power. In other 
words, the onsite and offsite electric power systems are classified as 
non-Class 1E systems and electric power is not needed (1) to achieve or 
maintain safe shutdown, (2) to assure specified acceptable fuel design 
limits and design conditions of the reactor coolant pressure boundary 
are not exceeded as a result of anticipated operational occurrences, or 
(3) to maintain core cooling, containment integrity, and other vital 
functions during postulated accidents. Further, the onsite and offsite 
power systems are not needed to permit functioning of structures, 
systems, and components important to safety. Therefore, NuScale Power 
was granted an exemption from GDC 17. The NRC's evaluation of NuScale 
Power's exemption request from the requirements of GDC 17 is documented 
in Section 8.1.5, ``Technical Evaluation for Exemptions,'' of the final 
safety evaluation report for Chapter 8, ``Electric Power.''

F. Accident Source Term Methodology

    The NRC reviewed NuScale Power's methods for developing accident 
source terms and performing accident radiological consequence analyses. 
As defined in Sec.  50.2, ``Definitions,'' a source term ``refers to 
the magnitude and mix of the radionuclides released from the fuel, 
expressed as fractions of the fission product inventory in the fuel, as 
well as their physical and chemical form, and the timing of their 
release.'' NuScale Power developed source terms for deterministic 
accidents for NuScale that are similar to those that have been used in 
safety and siting assessments for large light water reactors. The 
design-basis accidents for NuScale are the main steam line break 
outside containment, rod ejection accident, fuel handling accident, 
steam generator tube failure, and the failure of small lines carrying 
primary coolant outside containment.
    To address the source term regulatory requirements, NuScale Power 
submitted topical report TR-0915-17565, Revision 3, ``Accident Source 
Term Methodology,'' dated April 2019. The topical report proposes a 
methodology to develop a source term based on several severe accident 
scenarios that result in core damage, taken from the design 
probabilistic risk assessment. This source term is the surrogate 
radiological source term for a core damage event.
    The topical report also provides methods for determining radiation 
sources not developed from core damage scenarios for use in the 
evaluation of environmental qualification of equipment under Sec.  
50.49, ``Environmental qualification of electric equipment important to 
safety for nuclear power plants.'' Specifically, the report describes 
an iodine spike source term not involving core damage, which is a 
surrogate accident that bounds potential accidents with release of the 
reactor coolant into the containment vessel.

[[Page 3294]]

    The NRC staff submitted a related information paper to the 
Commission, SECY-19-0079, ``Staff Approach to Evaluate Accident Source 
Terms for the NuScale Power Design Certification Application,'' dated 
August 16, 2019, describing the regulatory and technical issues raised 
by unique aspects of NuScale Power's methodology and the staff's 
approach to reviewing topical report TR-0915-17565.
    The NRC's review and findings of topical report TR-0915-17565, 
Revision 3, are documented in the topical report final safety 
evaluation report issued on October 24, 2019. The approved version of 
topical report TR-0915-17565-NP-A, Revision 4, is discussed in the 
final safety evaluation report Section 12.2, ``Radiation Sources,'' 
Section 12.3, ``Radiation Protection Design Features,'' Section 3.11 
``Environmental Qualification of Mechanical and Electrical Equipment,'' 
Section 15.0.2, ``Review of Transient and Accident Analysis Methods,'' 
and Section 15.0.3, ``Radiological Consequences of Design Basis 
Accidents.'' The NRC found the accident source terms acceptable for the 
purposes described in each of the above safety evaluation report 
sections.

G. Boron Redistribution During Passive Cooling Modes

    The NRC evaluated the effects of boron volatility and 
redistribution during long term passive cooling. During this mode of 
operation, boron-free steam will enter the downcomer and containment, 
which can potentially challenge reactor core shutdown margin and could 
lead to a return to power. The NRC reviewed analyses provided by 
NuScale Power demonstrating that the reactor remains subcritical and 
that specified acceptable fuel design limits are not exceeded. The NRC 
evaluated the technical basis for NuScale Power's approach and 
conducted confirmatory calculations and independent assessments to 
determine its acceptability. The staff's review is primarily documented 
in Chapter 15, Section 15.0.5, ``Long Term Decay Heat and Residual Heat 
Removal,'' and Section 15.6.5, ``Loss of Coolant Accidents Resulting 
from Spectrum of Postulated Piping Breaks within the Reactor Coolant 
Pressure Boundary,'' of the safety evaluation report. Specifically, the 
staff concluded that the top of active fuel remains covered with 
acceptably low cladding temperatures and that for beginning-of-cycle 
and middle-of-cycle conditions, with no operator actions, the core 
remains subcritical. The potential for an end-of-cycle return to power 
is discussed in Section IV.D, ``Conformance with General Design 
Criterion 27, `Combined Reactivity Control Systems Capability,' '' of 
this document. In addition, Chapter 19, Section 19.1.4.6.4, ``Success 
Criteria, Accident Sequences, and Systems Analyses,'' of the safety 
evaluation report concludes that an operator error during recovery of 
the module from an uneven boron distribution scenario is unlikely to 
lead to core damage and is not a significant risk contributor.

H. Exemptions

    NuScale Power submitted a total of 17 requests for exemptions from 
the following regulations, including those discussed as part of the 
significant technical issues mentioned previously (see Table 1.14-1, 
``NuScale Design Certification Exemptions,'' in Chapter 1 of the final 
safety evaluation report):

1. Sec. Sec.  50.46a and 50.34(f)(2)(vi) (Reactor Coolant System 
Venting)
2. Sec.  50.44 (Combustible Gas Control)
3. Sec.  50.62(c)(1) (Reduction of Risk from Anticipated Transients 
Without Scram)
4. Appendix A to 10 CFR part 50, GDC 17, ``Electric Power Systems''; 
GDC 18, ``Inspection and Testing of Electric Power Systems''; and 
related provisions of GDC 34, ``Residual Heat removal''; GDC 35, 
``Emergency Core Cooling''; GDC 38, ``Containment Heat Removal''; GDC 
41, ``Containment Atmosphere Cleanup''; and GDC 44, ``Cooling Water'' 
(Electric Power Systems GDCs)
5. Appendix A to 10 CFR part 50, GDC 33, ``Reactor Coolant Makeup''
6. Sec.  50.54(m) (Control Room Staffing) (Alternative to meet the 
regulation)
7. Appendix A to 10 CFR part 50, GDC 52, ``Capability for Containment 
Leakage Rate Testing'' and Appendix J to 10 CFR part 50 (Type A 
testing)
8. Appendix A to 10 CFR part 50, GDC 40, ``Testing of Containment Heat 
Removal System''
9. Appendix A to 10 CFR part 50, GDC 55, ``Reactor Coolant Pressure 
Boundary Penetrating Containment,'' GDC 56, ``Primary Containment 
Isolation,'' and GDC 57, ``Closed Systems Isolation Valves'' 
(Containment Isolation)
10. Appendix K to 10 CFR part 50 (Emergency Core Cooling System 
Evaluation Models)
11. Sec.  50.34(f)(2)(xx) (Power Supplies for Pressurizer Relief 
Valves, Block Valves, and Level Indicators)
12. Sec.  50.34(f)(2)(xiii) (Pressurizer Heater Power Supplies)
13. Sec.  50.34(f)(2)(xiv)(E) (Containment Evacuation System Isolation)
14. Sec.  50.46 (Fuel Rod Cladding Material)
15. Appendix A to 10 CFR part 50, GDC 27, ``Combined Reactivity Control 
Systems Capability''
16. Sec.  50.34(f)(2)(viii) (Post-Accident Sampling)
17. Appendix A to 10 CFR part 50, GDC 19, ``Control Room''

    NRC's safety evaluation report for Chapter 1, ``Introduction and 
General Discussion,'' Section 1.14, ``Index of Exemptions,'' lists 
these exemption requests with the corresponding sections of the safety 
evaluation report where these exemption requests have been evaluated. 
The NRC granted each exemption request.

I. Differing Professional Opinion Related to Chapter 3 of NuScale

    On September 17, 2020, a Differing Professional Opinion (DPO) was 
submitted that raised concerns related to the seismic margin evaluation 
of the NuScale reactor building and its structural response during the 
review level earthquake. An ad-hoc review panel was formed and tasked 
to review the DPO. The review panel subsequently issued its report to 
the Director of the Office of Nuclear Reactor Regulation (NRR) on April 
19, 2021. On May 19, 2021, the Director of NRR issued a decision to the 
DPO submitter. For the reasons described in the decision, the Director 
of NRR agreed with the review panel's finding that the NuScale reactor 
building design was complete and acceptable for the purposes of a 
design certification application. On June 14, 2021, the DPO submitter 
appealed the DPO decision to the Executive Director for Operations 
(EDO).
    After consideration of the issues raised in the appeal, the EDO 
issued a decision on the DPO appeal on February 8, 2022. The EDO 
directed NRR to (1) document its evaluation of the stress averaging 
approach used in the NuScale design certification application, 
including, if necessary, updating the Final Safety Evaluation Report 
and assess whether there are any impacts to the standard design 
approval, and (2) evaluate and update guidance, or create knowledge 
management tools, on how to assess applications that use stress 
averaging for structural building design. On February 14, 2022, the DPO 
submitter responded to the EDO's DPO appeal decision. In this response, 
the submitter thanked the EDO for thoughtful consideration of the 
concerns raised and provided clarification regarding the applicability 
of the Probabilistic Risk Assessment-based seismic margin analysis to 
the reactor building. After reviewing and considering the submitter's 
response to

[[Page 3295]]

the DPO appeal decision, on March 15, 2022, the EDO directed the NRC 
staff to review and consider the totality of the information provided 
by the submitter when addressing the tasks mandated in the DPO appeal 
decision.
    In response to the EDO tasking, on May 13, 2022, the Director of 
NRR issued a memo to the EDO (``Response to DPO Tasking'') discussing 
the staff's review of the items described in the tasking, documenting 
the staff's evaluation of the approach used in the NuScale design 
certification, and detailing the staff's assessment of existing related 
structural analysis guidance (ADAMS Accession No. ML22062A007). The 
Director of NRR concluded that the staff sufficiently assessed the 
evaluation of the demand (force/moment) averaging approach used in the 
NuScale DCA; justified the acceptability to conclude that there are no 
impacts to the NuScale standard design approval issued in September 
2020; determined that an update or supplement to the final safety 
evaluation report for the NuScale DCA is not necessary; and found that 
the existing review guidance is sufficient to review and evaluate an 
applicant's structural analysis/design. Details on the EDO's decision 
on the DPO appeal and related correspondence, and the Response to DPO 
Tasking are found in the information package for DPO-2020-004 (ADAMS 
Accession No. ML22122A116).
    The NRC staff's assessment of NuScale's use of the demand (force/
moment) averaging approach is documented in the Response to DPO 
Tasking. The Response to DPO Tasking elaborates on the reasons for, but 
does not change, the conclusion in the final safety evaluation report. 
Based on this assessment, the NRC concludes that the use of the demand 
(force/moment) averaging approach is acceptable, as stated in the final 
safety evaluation report.

V. Discussion

Final Safety Evaluation Report

    NuScale Power submitted the final revision of the NuScale DCA, 
Revision 5, in July 2020 (ADAMS Accession No. ML20225A071). In August 
2020, the NRC issued a final safety evaluation report after the 
Advisory Committee on Reactor Safeguards (ACRS) performed its final 
independent review and issued its July 29, 2020, letter to the 
Commission on its findings and recommendations. The final safety 
evaluation report is a collection of reports written by the NRC 
documenting the safety findings from its review of the standard design 
application, and it reflects all changes resulting from interactions 
with the ACRS as well as changes in the final version of the DCA. The 
final safety evaluation report, as elaborated on by the Response to DPO 
Tasking, reflects that NuScale Power has resolved all technical and 
safety issues with the exception of the three issues discussed 
previously. As noted above, the Response to DPO Tasking elaborates on 
the reasons for, but does not change, the conclusion in the final 
safety evaluation report that NuScale's use of the demand (force/
moment) averaging approach is acceptable as a realistic engineering 
practice.
    In addition, the final safety evaluation report describes the 
portions of the design that are not receiving finality in this rule 
and, therefore, are not part of the certified design. The final safety 
evaluation report also includes an index of all NRC requests for 
additional information, a chronology of all documents related to the 
NuScale DCA review, and summaries of public meetings and audits.

NuScale Design Certification Final Rule

    This section describes the purpose and key aspects of each section 
of this NuScale design certification final rule. All section and 
paragraph references are to the provisions being added as appendix G to 
10 CFR part 52, unless otherwise noted. The NRC has modeled this 
NuScale design certification final rule on existing design 
certification rules, with certain modifications where necessary to 
account for differences in the design documentation, design features, 
and environmental assessment (including severe accident mitigation 
design alternatives). As a result, design certification rules are 
standardized to the extent practical.

A. Introduction (Section I)

    The purpose of Section I of appendix G to 10 CFR part 52 is to 
identify the standard design that is approved by this design 
certification final rule and the applicant for certification of the 
standard design. Identification of the design certification applicant 
is necessary to implement appendix G to 10 CFR part 52 for two reasons. 
First, the implementation of Sec.  52.63(c) depends on whether an 
applicant contracts with the design certification applicant to obtain 
the generic DCD and supporting design information. If a COL applicant 
does not use the design certification applicant to provide the design 
information and instead uses an alternate vendor, then the COL 
applicant must meet the requirements in Sec.  52.73. Second, paragraph 
X.A.1 requires that the identified design certification applicant 
maintain the generic DCD throughout the time that appendix G to 10 CFR 
part 52 may be referenced.

B. Definitions (Section II)

    The purpose of Section II of appendix G to 10 CFR part 52 is to 
define specific terminology with respect to this design certification 
final rule. During development of the first two design certification 
rules, the NRC decided that there would be both generic DCDs maintained 
by the NRC and the design certification applicant, as well as 
individual plant-specific DCDs maintained by each applicant or licensee 
that references a 10 CFR part 52 appendix. This distinction is 
necessary in order to specify the relevant plant-specific requirements 
to applicants and licensees referencing appendix G to 10 CFR part 52.
    In order to facilitate the maintenance of the generic DCDs, the NRC 
requires that applicants for a standard design certification update 
their application to include an electronic copy of the final version of 
the DCD. The final version incorporates all amendments to the DCA 
submitted since the original application and any changes directed by 
the NRC as a result of its review of the original DCA or as a result of 
public comments. This final version is then incorporated by reference 
in the design certification rule. Once incorporated by reference, the 
final version becomes the ``generic DCD,'' which will be maintained by 
the design certification applicant and the NRC and updated as needed to 
include any generic changes made after this design certification 
rulemaking. These changes would occur as the result of generic 
rulemaking by the NRC, under the change criteria in Section VIII of 
appendix G to 10 CFR part 52.
    The NRC also requires each applicant and licensee referencing 
appendix G to 10 CFR part 52 to submit and maintain a plant-specific 
DCD as part of the COL final safety analysis report. The plant-specific 
DCD must either include or incorporate by reference the information in 
the generic DCD. The COL licensee is required to maintain the plant-
specific DCD, updating it as necessary to reflect the generic changes 
to the DCD that the NRC may adopt through rulemaking, plant-specific 
departures from the generic DCD that the NRC imposes on the licensee by 
order, and any plant-specific departures that the licensee chooses to 
make in accordance with the relevant processes in Section VIII of 
appendix G to 10 CFR part 52. A COL applicant will also have to include 
considerations for a multi-unit site in

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the plant-specific DCD that were not previously evaluated as part of 
the design certification rule, e.g., construction impacts on operating 
units. Therefore, the plant-specific DCD functions like an updated 
final safety analysis report because it would provide the most complete 
and accurate information on a plant's design basis for that part of the 
plant that would be within the scope of appendix G to 10 CFR part 52.
    The NRC is treating the technical specifications in Part 4, 
``Technical Specifications,'' of the DCA as a special category of 
information and designating them as generic technical specifications in 
order to facilitate the special treatment of this information under 
appendix G to 10 CFR part 52. A COL applicant must submit plant-
specific technical specifications that consist of the generic technical 
specifications, which may be modified as specified in paragraph VIII.C, 
and the remaining site-specific information needed to complete the 
technical specifications. The final safety analysis report that is 
required by Sec.  52.79 will consist of the plant-specific DCD, the 
site-specific final safety analysis report, and the plant-specific 
technical specifications.
    The terms Tier 1, Tier 2, and COL items (license information) are 
defined in appendix G to 10 CFR part 52 because these concepts were not 
envisioned when 10 CFR part 52 was developed. The design certification 
applicants and the NRC use these terms in implementing a two-tiered 
rule structure (the DCD is divided into Tier 1 and Tier 2 to support 
the rule structure) that was proposed by representatives of the nuclear 
industry after publication of 10 CFR part 52. The Commission approved 
the use of the two-tiered rule structure in its staff requirements 
memorandum (SRM), dated February 15, 1991, on SRM-SECY-90-377, 
``Requirements for Design Certification under 10 CFR part 52,'' dated 
November 8, 1990.
    Tier 1 information means the portion of the design-related 
information contained in the generic DCD that is approved and certified 
by this appendix. Tier 2 information means the portion of the design-
related information contained in the generic DCD that is approved but 
not certified by this appendix. The change process for Tier 2 
information is similar, but not identical to, the change process set 
forth in Sec.  50.59. The regulations in Sec.  50.59 describe when a 
licensee may make changes to a plant as described in its final safety 
analysis report without a license amendment. Because of some 
differences in how the change control requirements are structured in 
the design certification rules, certain definitions contained in Sec.  
50.59 are not applicable to 10 CFR part 52 and are not being included 
in this final rule. The NRC is including a definition for ``Departure 
from a method of evaluation'' in paragraph II.F of appendix G to 10 CFR 
part 52, so that the eight criteria in paragraph VIII.B.5.b will be 
implemented for new reactors as intended.

C. Scope and Contents (Section III)

    The purpose of Section III of appendix G to 10 CFR part 52 is to 
describe and define the scope and content of this design certification, 
explain how to obtain a copy of the generic DCD, identify requirements 
for incorporation by reference of the design certification rule, and 
set forth how documentation discrepancies or inconsistencies are to be 
resolved.
    Paragraph III.A is the required statement of the Office of the 
Federal Register for approval of the incorporation by reference of the 
NuScale DCD, Revision 5. In addition, this paragraph provides the 
information on how to obtain a copy of the DCD. Unlike previous design 
certifications, the documents submitted to the NRC by NuScale Power did 
not use the title ``Design Control Document;'' they used the title 
``Design Certification Application'' instead.
    Paragraph III.B is the requirement for applicants and licensees 
referencing appendix G to 10 CFR part 52. The legal effect of 
incorporation by reference is that the incorporated material has the 
same legal status as if it were published in the Code of Federal 
Regulations. This material, like any other properly issued regulation, 
has the force and effect of law. Tier 1 and Tier 2 information 
(including the technical and topical reports referenced in the DCD Tier 
2, Chapter 1) and generic technical specifications have been combined 
into a single document called the generic DCD in order to effectively 
control this information and facilitate its incorporation by reference 
into the rule. In addition, paragraph III.B clarifies that the 
conceptual design information and NuScale Power's evaluation of severe 
accident mitigation design alternatives are not considered to be part 
of appendix G to 10 CFR part 52. As provided by Sec.  52.47(a)(24), 
these conceptual designs are not part of appendix G to 10 CFR part 52 
and, therefore, are not applicable to an application that references 
appendix G to 10 CFR part 52. Therefore, an applicant would not be 
required to conform to the conceptual design information that was 
provided by the design certification applicant. The conceptual design 
information, which consists of site-specific design features, was 
required to facilitate the design certification review. Similarly, the 
severe accident mitigation design alternatives were required to 
facilitate the environmental assessment.
    Paragraphs III.C and III.D set forth the manner by which potential 
conflicts are to be resolved and identify the controlling document. 
Paragraph III.C establishes the Tier 1 description in the DCD as 
controlling in the event of an inconsistency between the Tier 1 and 
Tier 2 information in the DCD. Paragraph III.D establishes the generic 
DCD as the controlling document in the event of an inconsistency 
between the DCD and the final safety evaluation report for the 
certified standard design.
    Paragraph III.E makes it clear that design activities outside the 
scope of the design certification may be performed using actual site 
characteristics. This provision applies to site-specific portions of 
the plant, such as the administration building.

D. Additional Requirements and Restrictions (Section IV)

    Section IV of appendix G to 10 CFR part 52 sets forth additional 
requirements and restrictions imposed upon an applicant who references 
appendix G to 10 CFR part 52.
    Paragraph IV.A sets forth the information requirements for COL 
applicants and distinguishes between information and documents that 
must be included in the application or the DCD and those which may be 
incorporated by reference. Any incorporation by reference in the 
application should be clear and should specify the title, date, edition 
or version of a document, the page number(s), and table(s) containing 
the relevant information to be incorporated. The legal effect of such 
an incorporation by reference into the application is that appendix G 
to 10 CFR part 52 would be legally binding on the applicant or 
licensee.
    In paragraph IV.B the NRC reserves the right to determine how 
appendix G to 10 CFR part 52 may be referenced under 10 CFR part 50. 
This determination may occur in the context of a subsequent rulemaking 
modifying 10 CFR part 52 or this design certification rule, or on a 
case-by-case basis in the context of a specific application for a 10 
CFR part 50 construction permit or operating license. This provision is 
necessary because the previous design certification rules were not

[[Page 3297]]

implemented in the manner that was originally envisioned at the time 
that 10 CFR part 52 was issued. The NRC's concern is with the manner by 
which the inspections, tests, analyses, and acceptance criteria (ITAAC) 
were developed and the lack of experience with design certifications in 
a licensing proceeding. Therefore, it is appropriate that the NRC 
retain some discretion regarding the manner by which appendix G to 10 
CFR part 52 could be referenced in a 10 CFR part 50 licensing 
proceeding.
    In paragraph IV.C, the NRC lists design-specific regulations that 
apply to licenses that reference this appendix.

E. Applicable Regulations (Section V)

    The purpose of Section V of appendix G to 10 CFR part 52 is to 
specify the regulations that were applicable and in effect at the time 
this design certification was approved. These regulations consist of 
the technically relevant regulations identified in paragraph V.A, 
except for the regulations in paragraph V.B that would not be 
applicable to this certified design.

F. Issue Resolution (Section VI)

    The purpose of Section VI of appendix G to 10 CFR part 52 is to 
identify the scope of issues that are resolved by the NRC through this 
final rule and, therefore, are ``matters resolved'' within the meaning 
and intent of Sec.  52.63(a)(5). The section is divided into five 
parts: paragraph VI.A identifies the NRC's safety findings in adopting 
appendix G to 10 CFR part 52, paragraph VI.B identifies the scope and 
nature of issues that are resolved by this final rule, paragraph VI.C 
identifies issues that are not resolved by this final rule, and 
paragraph VI.D identifies the issue finality restrictions applicable to 
the NRC with respect to appendix G to 10 CFR part 52.
    Paragraph VI.A describes the nature of the NRC's findings in 
general terms and makes the findings required by Sec.  52.54 for the 
NRC's approval of this design certification final rule.
    Paragraph VI.B sets forth the scope of issues that may not be 
challenged as a matter of right in subsequent proceedings. The 
introductory phrase of paragraph VI.B clarifies that issue resolution, 
as described in the remainder of the paragraph, extends to the 
delineated NRC proceedings referencing appendix G to 10 CFR part 52. 
The remainder of paragraph VI.B describes the categories of information 
for which there is issue resolution.
    Paragraph VI.C reserves the right of the NRC to impose operational 
requirements on applicants that reference appendix G to 10 CFR part 52. 
This provision reflects the fact that only some operational 
requirements, including portions of the generic technical specification 
in Chapter 16 of the DCD, were completely or comprehensively reviewed 
by the NRC in this design certification final rule proceeding. The NRC 
notes that operational requirements may be imposed on licensees 
referencing this design certification through the inclusion of license 
conditions in the license or inclusion of a description of the 
operational requirement in the plant-specific final safety analysis 
report.\1\ The NRC's choice of the regulatory vehicle for imposing the 
operational requirements will depend upon, among other things, (1) 
whether the development and/or implementation of these requirements 
must occur prior to either the issuance of the COL or the Commission 
finding under Sec.  52.103(g), and (2) the nature of the change 
controls that are appropriate given the regulatory, safety, and 
security significance of each operational requirement.
---------------------------------------------------------------------------

    \1\ Certain activities ordinarily conducted following fuel load 
and, therefore, considered ``operational requirements,'' but which 
may be relied upon to support a Commission finding under Sec.  
52.103(g), may themselves be the subject of ITAAC to ensure their 
implementation prior to the Sec.  52.103(g) finding.
---------------------------------------------------------------------------

    Also, paragraph VI.C allows the NRC to impose future operational 
requirements (distinct from design matters) on applicants who reference 
this design certification. License conditions for portions of the plant 
within the scope of this design certification (e.g., startup and power 
ascension testing) are not restricted by Sec.  52.63. The requirement 
to perform these testing programs is contained in the Tier 1 
information. However, ITAAC cannot be specified for these subjects 
because the matters to be addressed in these license conditions cannot 
be verified prior to fuel load and operation when the ITAAC are 
satisfied. In the absence of detailed design information to evaluate 
the need for and develop specific post-fuel load verifications for 
these matters, the NRC is reserving the right to impose, at the time of 
COL issuance, license conditions addressing post-fuel load verification 
activities for portions of the plant within the scope of this design 
certification.
    Paragraph VI.D reiterates the restrictions (contained in Section 
VIII of appendix G to 10 CFR part 52) placed upon the NRC when ordering 
generic or plant-specific modifications, changes, or additions to 
structures, systems, and components, design features, design criteria, 
and ITAAC within the scope of the certified design.
    Paragraph VI.E provides that the NRC will specify at an appropriate 
time the procedures on how to obtain access to sensitive unclassified 
and non-safeguards information (SUNSI) and safeguards information (SGI) 
for the NuScale design certification rule. Access to such information 
would be for the sole purpose of requesting or participating in certain 
specified hearings, such as hearings required by Sec.  52.85 or an 
adjudicatory hearing. For proceedings where the notice of hearing was 
published before the effective date of the final rule, the Commission's 
order governing access to SUNSI and SGI shall be used to govern access 
to such information within the scope of the rulemaking. For proceedings 
in which the notice of hearing or opportunity for hearing is published 
after the effective date of the final rule, paragraph VI.E applies and 
governs access to SUNSI and SGI.

G. Duration of This Appendix (Section VII)

    The purpose of Section VII of appendix G to 10 CFR part 52 is, in 
part, to specify the period during which this design certification may 
be referenced by an applicant, under Sec.  52.55, and the period it 
will remain valid when the design certification is referenced. For 
example, if an application references this design certification during 
the 15-year period, then the design certification would be effective 
until the application is withdrawn or the license issued on that 
application expires. The NRC intends for appendix G to 10 CFR part 52 
to remain valid for the life of any license that references the design 
certification to achieve the benefits of standardization and licensing 
stability. This means that changes to, or plant-specific departures 
from, information in the plant-specific DCD must be made under the 
change processes in Section VIII for the life of the plant.

H. Processes for Changes and Departures (Section VIII)

    The purpose of Section VIII of appendix G to 10 CFR part 52 is to 
set forth the processes for generic changes to, or plant-specific 
departures (including exemptions) from, the DCD. The NRC adopted this 
restrictive change process in order to achieve a more stable licensing 
process for applicants and licensees that reference design 
certification rules. Section VIII is divided into three paragraphs, 
which correspond to Tier 1, Tier 2, and operational requirements.

[[Page 3298]]

    Generic changes (called ``modifications'' in Sec.  52.63(a)(3)) 
must be accomplished by rulemaking because the intended subject of the 
change is this design certification rule itself, as is contemplated by 
Sec.  52.63(a)(1). Consistent with Sec.  52.63(a)(3), any generic 
rulemaking changes are applicable to all plants, absent circumstances 
which render the change technically irrelevant. By contrast, plant-
specific departures could be required by either an order to one or more 
applicants or licensees; or an applicant or licensee-initiated 
departure applicable only to that applicant's or licensee's plant(s), 
similar to a Sec.  50.59 departure or an exemption. Because these 
plant-specific departures will result in a DCD that is unique for that 
plant, Section X requires an applicant or licensee to maintain a plant-
specific DCD. For purposes of brevity, the following discussion refers 
to the processes for both generic changes and plant-specific departures 
as ``change processes.'' Section VIII refers to an exemption from one 
or more requirements of this appendix and addresses the criteria for 
granting an exemption. The NRC cautions that when the exemption 
involves an underlying substantive requirement (i.e., a requirement 
outside this appendix), then the applicant or licensee requesting the 
exemption must demonstrate that an exemption from the underlying 
applicable requirement meets the criteria of Sec. Sec.  52.7 and 50.12.
    For the NuScale review, the staff followed the approach described 
in SECY-17-0075, ``Planned Improvements in Design Certification Tiered 
Information Designations,'' dated July 24, 2017, to evaluate the 
applicant's designation of information as Tier 1 or Tier 2 information. 
Unlike some of the prior DCAs, this application did not contain any 
Tier 2* information. As described in SECY-17-0075, prior design 
certification rules in 10 CFR part 52, appendices A through E, 
information contained in the DCD was divided into three designations: 
Tier 1, Tier 2, and Tier 2*. Tier 1 information is the portion of 
design-related information in the generic DCD that the Commission 
approves in the 10 CFR part 52 design certification rule appendices. To 
change Tier 1 information, NRC approval by rulemaking or approval of an 
exemption from the certified design rule is required. Tier 2 
information is also approved by the Commission in the 10 CFR part 52 
design certification rule appendices, but it is not certified and 
licensees who reference the design can change this information using 
the process outlined in Section VIII of the appendices. This change 
process is similar to that in Sec.  50.59 and is generally referred to 
as the ``50.59-like'' process. If the criteria in Section VIII are met, 
a licensee can change Tier 2 information without prior NRC approval.
    As mentioned in the previous paragraph, the NRC created a third 
category, Tier 2*, in other design certification rules. This third 
category was created to address industry requests to minimize the scope 
of Tier 1 information and provide greater flexibility for making 
changes. Unlike Tier 2 information, all changes to Tier 2* information 
require a license amendment, but unlike Tier 1 information, no 
exemption is required. In those rules, Tier 2* information has the same 
safety significance as Tier 1 information but is part of the Tier 2 
section of the DCD to afford more flexibility for licensees to change 
this type of information.
    The applicant did not designate or categorize any Tier 2* 
information in the NuScale DCA. The NRC evaluated the Tier 2 
information to determine whether any of that information should require 
NRC approval before it is changed. If the NRC had identified any such 
information in Tier 2, then the NRC would have requested that the 
applicant revise the application to categorize that information as Tier 
1 or Tier 2*. The NRC did not identify any information in Tier 2 that 
should be categorized as Tier 2*. Because neither the applicant nor the 
NRC have designated any information in the DCD as Tier 2*, that 
designation and related requirements are not being used in this design 
certification rule.
Tier 1 Information
    Paragraph A of Section VIII describes the change process for 
changes to Tier 1 information that are accomplished by rulemakings that 
amend the generic DCD and are governed by the standards in Sec.  
52.63(a)(1). A generic change under Sec.  52.63(a)(1) will not be made 
to a certified design while it is in effect unless the change: (1) is 
necessary for compliance with NRC regulations applicable and in effect 
at the time the certification was issued; (2) is necessary to provide 
adequate protection of the public health and safety or common defense 
and security; (3) reduces unnecessary regulatory burden and maintains 
protection to public health and safety and common defense and security; 
(4) provides the detailed design information necessary to resolve 
select design acceptance criteria; (5) corrects material errors in the 
certification information; (6) substantially increases overall safety, 
reliability, or security of a facility and the costs of the change are 
justified; or (7) contributes to increased standardization of the 
certification information. The rulemakings must provide for notice and 
opportunity for public comment on the proposed change under Sec.  
52.63(a)(2). The NRC will give consideration as to whether the benefits 
justify the costs for plants that are already licensed or for which an 
application for a permit or license is under consideration.
    Departures from Tier 1 may occur in two ways: (1) the NRC may order 
a licensee to depart from Tier 1, as provided in paragraph VIII.A.3; or 
(2) an applicant or licensee may request an exemption from Tier 1, as 
addressed in paragraph VIII.A.4. If the NRC seeks to order a licensee 
to depart from Tier 1, paragraph VIII.A.3 would require that the NRC 
find both that the departure is necessary for adequate protection or 
for compliance and that special circumstances are present. Paragraph 
VIII.A.4 provides that exemptions from Tier 1 requested by an applicant 
or licensee are governed by the requirements of Sec. Sec.  52.63(b)(1) 
and 52.98(f), which provide an opportunity for a hearing. In addition, 
the NRC would not grant requests for exemptions that will result in a 
significant decrease in the level of safety otherwise provided by the 
design.
Tier 2 Information
    Paragraph B of Section VIII describes the change processes for the 
Tier 2 information, which have the same elements as the Tier 1 change 
process, but some of the standards for plant-specific orders and 
exemptions would be different. Generic Tier 2 changes would be 
accomplished by rulemaking that would amend the generic DCD and would 
be governed by the standards in Sec.  52.63(a)(1). A generic change 
under Sec.  52.63(a)(1) would not be made to a certified design while 
it is in effect unless the change: (1) is necessary for compliance with 
NRC regulations that were applicable and in effect at the time the 
certification was issued; (2) is necessary to provide adequate 
protection of the public health and safety or common defense and 
security; (3) reduces unnecessary regulatory burden and maintains 
protection to public health and safety and common defense and security; 
(4) provides the detailed design information necessary to resolve 
select design acceptance criteria; (5) corrects material errors in the 
certification information; (6)

[[Page 3299]]

substantially increases overall safety, reliability, or security of a 
facility and the costs of the change are justified; or (7) contributes 
to increased standardization of the certification information.
    Departures from Tier 2 would occur in four ways: (1) the NRC may 
order a plant-specific departure, as set forth in paragraph VIII.B.3; 
(2) an applicant or licensee may request an exemption from a Tier 2 
requirement as set forth in paragraph VIII.B.4; (3) a licensee may make 
a departure without prior NRC approval under paragraph VIII.B.5; or (4) 
the licensee may request NRC approval for proposed departures that do 
not meet the requirements in paragraph VIII.B.5 as provided in 
paragraph VIII.B.5.e.
    Similar to ordered Tier 1 departures and generic Tier 2 changes, 
ordered Tier 2 departures could not be imposed except when necessary, 
either to bring the certification into compliance with the NRC's 
regulations applicable and in effect at the time of approval of the 
design certification or to ensure adequate protection of the public 
health and safety or common defense and security, as set forth in 
paragraph VIII.B.3. However, unlike Tier 1 departures, the Commission 
would not have to consider whether the special circumstances for the 
Tier 2 departures would outweigh any decrease in safety that may result 
from the reduction in standardization caused by the plant-specific 
order, as required by Sec.  52.63(a)(4). The NRC has determined that it 
is not necessary to impose an additional limitation for standardization 
similar to that imposed on Tier 1 departures by Sec.  52.63(a)(4) and 
(b)(1) because it would unnecessarily restrict the flexibility of 
applicants and licensees with respect to Tier 2 information.
    An applicant or licensee may request an exemption from Tier 2 
information as set forth in paragraph VIII.B.4. The applicant or 
licensee would have to demonstrate that the exemption complies with one 
of the special circumstances in regulations governing specific 
exemptions in Sec.  50.12(a). In addition, the NRC would not grant 
requests for exemptions that will result in a significant decrease in 
the level of safety otherwise provided by the design. However, unlike 
Tier 1 changes, the special circumstances for the exemption do not have 
to outweigh any decrease in safety that may result from the reduction 
in standardization caused by the exemption. If the exemption is 
requested by an applicant for a license, the exemption would be subject 
to litigation in the same manner as other issues in the licensing 
hearing, consistent with Sec.  52.63(b)(1). If the exemption is 
requested by a licensee, then the exemption would be subject to 
litigation in the same manner as a license amendment.
    Paragraph VIII.B.5 allows an applicant or licensee to depart from 
Tier 2 information, without prior NRC approval, if it does not involve 
a change to, or departure from, Tier 1 information, technical 
specification, or does not require a license amendment under paragraphs 
VIII.B.5.b or c. The technical specifications referred to in VIII.B.5.a 
of this paragraph are the technical specifications in Chapter 16 of the 
generic DCD, including bases, for departures made prior to the issuance 
of the COL. After the issuance of the COL, the plant-specific technical 
specifications would be controlling under paragraph VIII.B.5. The 
requirement for a license amendment in paragraph VIII.B.5.b is similar 
to the requirement in Sec.  50.59 and applies to all of the information 
in Tier 2 except for the information that resolves the severe accident 
issues or the information required by Sec.  52.47(a)(28) to address 
aircraft impacts.
    Paragraph VIII.B.5.d addresses information described in the DCD to 
address aircraft impacts, in accordance with Sec.  52.47(a)(28). Under 
Sec.  52.47(a)(28), applicants are required to include the information 
required by Sec.  50.150(b) in their DCD. An applicant or licensee who 
changes this information is required to consider the effect of the 
changed design feature or functional capability on the original 
aircraft impact assessment required by Sec.  50.150(a). The applicant 
or licensee is also required to describe in the plant-specific DCD how 
the modified design features and functional capabilities continue to 
meet the assessment requirements in Sec.  50.150(a)(1). Submittal of 
this updated information is governed by the reporting requirements in 
Section X.B.
    During an ongoing adjudicatory proceeding (e.g., for issuance of a 
COL), a party who believes that an applicant or licensee has not 
complied with paragraph VIII.B.5 when departing from Tier 2 information 
may petition to admit such a contention into the proceeding under 
paragraph VIII.B.5.g. As set forth in paragraph VIII.B.5.g, the 
petition would have to comply with the NRC's hearing requirements at 
Sec.  2.309 and show that the departure does not comply with paragraph 
VIII.B.5. If on the basis of the petition and any responses thereto, 
the presiding officer in the proceeding determines that the required 
showing has been made, the matter would be certified to the Commission 
for its final determination. In the absence of a proceeding, assertions 
of nonconformance with paragraph VIII.B.5 requirements applicable to 
Tier 2 departures would be treated as petitions for enforcement action 
under Sec.  2.206.
Operational Requirements
    The change process for technical specifications and other 
operational requirements that were reviewed and approved in the design 
certification rule is set forth in Section VIII, paragraph C. The key 
to using the change processes described in Section VIII is to determine 
if the proposed change or departure would require a change to a design 
feature described in the generic DCD. If a design change is required, 
then the appropriate change process in paragraph VIII.A or VIII.B would 
apply. However, if a proposed change to the technical specifications or 
other operational requirements does not require a change to a design 
feature in the generic DCD, then paragraph VIII.C would apply. This 
change process has elements similar to the Tier 1 and Tier 2 change 
processes in paragraphs VIII.A and VIII.B, but with significantly 
different change standards. Because of the different finality status 
for technical specifications and other operational requirements, the 
NRC designated a special category of information, consisting of the 
technical specifications and other operational requirements, with its 
own change process in paragraph VIII.C. The language in paragraph 
VIII.C also distinguishes between generic (Chapter 16 of the DCD) and 
plant-specific technical specifications to account for the different 
treatment and finality consistent with technical specifications before 
and after a license is issued.
    The process in paragraph VIII.C.1 for making generic changes to the 
generic technical specifications or other operational requirements in 
the generic DCD is accomplished by rulemaking and governed by the 
backfit standards in Sec.  50.109. The determination of whether the 
generic technical specifications and other operational requirements 
were completely reviewed and approved in the design certification rule 
is based upon the extent to which the NRC reached a safety conclusion 
in the final safety evaluation report on this matter. If a technical 
specification or operational requirement was completely reviewed and 
finalized in the design certification rule, then the requirement of 
Sec.  50.109 would apply because a position was taken on that safety 
matter. Generic changes made under paragraph

[[Page 3300]]

VIII.C.1 would be applicable to all applicants or licensees (refer to 
paragraph VIII.C.2), unless the change is irrelevant because of a 
plant-specific departure.
    Some generic technical specifications contain values in brackets [ 
]. The brackets are placeholders indicating that the NRC's review is 
not complete and represent a requirement that an applicant for a COL 
referencing appendix G to 10 CFR part 52 must replace the values in 
brackets with final plant-specific values (refer to guidance provided 
in Regulatory Guide 1.206, Revision 1, ``Applications for Nuclear Power 
Plants,'' dated October 2018). The values in brackets are neither part 
of the design certification rule nor are they binding. Therefore, the 
replacement of bracketed values with final plant-specific values does 
not require an exemption from the generic technical specifications.
    Plant-specific departures may occur by either an order under 
paragraph VIII.C.3 or an applicant's exemption request under paragraph 
VIII.C.4. The basis for determining if the technical specification or 
operational requirement was completely reviewed and approved for these 
processes would be the same as for paragraph VIII.C.1 previously 
discussed. If the technical specification or operational requirement 
was comprehensively reviewed and finalized in the design certification 
rule, then the NRC must demonstrate that special circumstances are 
present before ordering a plant-specific departure. If not, there would 
be no restriction on plant-specific changes to the technical 
specifications or operational requirements, prior to the issuance of a 
license, provided a design change is not required. Although the generic 
technical specifications were reviewed and approved by the NRC in 
support of the design certification review, the NRC intends to consider 
the lessons learned from subsequent operating experience during its 
licensing review of the plant-specific technical specifications. The 
process for petitioning to intervene on a technical specification or 
operational requirement contained in paragraph VIII.C.5 is similar to 
other issues in a licensing hearing, except that the petitioner must 
also demonstrate why special circumstances are present pursuant to 
Sec.  2.335.
    Paragraph VIII.C.6 states that the generic technical specifications 
would have no further effect on the plant-specific technical 
specifications after the issuance of a license that references this 
appendix and the change process. After a license is issued, the bases 
for the plant-specific technical specification would be controlled by 
the bases change provision set forth in the administrative controls 
section of the plant-specific technical specifications.

I. [RESERVED] (Section IX)

    This section is reserved for future use. The matters discussed in 
this section of earlier design certification rules--inspections, tests, 
analyses, and acceptance criteria--are now addressed in the substantive 
provisions of 10 CFR part 52. Accordingly, there is no need to repeat 
these regulatory provisions in the NuScale design certification rule. 
However, this section is being reserved to maintain consistent section 
numbering with other design certification rules.

J. Records and Reporting (Section X)

    The purpose of Section X of appendix G to 10 CFR part 52 is to set 
forth the requirements that will apply to maintaining records of 
changes to and departures from the generic DCD, which are to be 
reflected in the plant-specific DCD. Section X also sets forth the 
requirements for submitting reports (including updates to the plant-
specific DCD) to the NRC. This section of appendix G to 10 CFR part 52 
is similar to the requirements for records and reports in 10 CFR part 
50, except for minor differences in information collection and 
reporting requirements.
    Paragraph X.A.1 requires that a generic DCD including referenced 
SUNSI and SGI be maintained by the applicant for this final rule. The 
generic DCD concept was developed, in part, to meet the requirements 
for incorporation by reference, including public availability of 
documents incorporated by reference. However, the SUNSI and SGI could 
not be included in the generic DCD because they are not publicly 
available. Nonetheless, the SUNSI and SGI were reviewed by the NRC and, 
as stated in paragraph VI.B.2, the NRC would consider the information 
to be resolved within the meaning of Sec.  52.63(a)(5). Because this 
information, or its equivalent, is not in the generic DCD, it is 
required to be provided by an applicant for a license referencing 
appendix G to 10 CFR part 52. Only the generic DCD is identified and 
incorporated by reference by this final rule. The generic DCD and the 
NRC approved version of the SUNSI and SGI must be maintained by the 
applicant (NuScale Power) for the period of time that appendix G to 10 
CFR part 52 may be referenced.
    Paragraphs X.A.2 and X.A.3 place recordkeeping requirements on the 
applicant or licensee that reference this design certification so that 
its plant-specific DCD accurately reflects both generic changes to the 
generic DCD and plant-specific departures made under Section VIII. The 
term ``plant-specific'' is used in paragraph X.A.2 and other sections 
of appendix G to 10 CFR part 52 to distinguish between the generic DCD 
that this final rule incorporates by reference into appendix G to 10 
CFR part 52, and the plant-specific DCD that the COL applicant is 
required to submit under paragraph IV.A. The requirement to maintain 
changes to the generic DCD is explicitly stated to ensure that these 
changes are not only reflected in the generic DCD, which will be 
maintained by the applicant for the design certification, but also in 
the plant-specific DCD. Therefore, records of generic changes to the 
DCD will be required to be maintained by both entities to ensure that 
both entities have up-to-date DCDs.
    Paragraph X.A.4.a requires the design certification rule applicant 
to maintain a copy of the aircraft impact assessment analysis for the 
term of the certification and any renewal. This provision, which is 
consistent with Sec.  50.150(c)(3), would facilitate any NRC 
inspections of the assessment that the NRC decides to conduct. 
Similarly, paragraph X.A.4.b requires an applicant or licensee who 
references appendix G to 10 CFR part 52 to maintain a copy of the 
aircraft impact assessment performed to comply with the requirements of 
Sec.  50.150(a) throughout the pendency of the application and for the 
term of the license and any renewal. This provision is consistent with 
Sec.  50.150(c)(4). For all applicants and licensees, the supporting 
documentation retained should describe the methodology used in 
performing the assessment, including the identification of potential 
design features and functional capabilities to show that the acceptance 
criteria in Sec.  50.150(a)(1) will be met.
    Paragraph X.A does not place recordkeeping requirements on site 
specific information that is outside the scope of this rule. As 
discussed in paragraph V.B of this document, the final safety analysis 
report required by Sec.  52.79 will contain the plant-specific DCD and 
the site-specific information for a facility that references this rule. 
The phrase ``site specific portion of the final safety analysis 
report'' in paragraph X.B.3.c refers to the information that is 
contained in the final safety analysis report for a facility (required 
by Sec.  52.79), but is not part of the plant-specific DCD (required by 
paragraph IV.A). Therefore, this final rule does not require that 
duplicate documentation be maintained by an applicant or licensee that 
references this

[[Page 3301]]

rule because the plant-specific DCD is part of the final safety 
analysis report for the facility.
    Paragraph X.B.1 requires applicants or licensees that reference 
this rule to submit reports that describe departures from the DCD and 
include a summary of the written evaluations. The requirement for the 
written evaluations is set forth in paragraph X.A.3. The frequency of 
the report submittals is set forth in paragraph X.B.3. The requirement 
for submitting a summary of the evaluations is similar to the 
requirement in Sec.  50.59(d)(2).
    Paragraph X.B.2 requires applicants or licensees that reference 
this rule to submit updates to the DCD, which include both generic 
changes and plant-specific departures, as set forth in paragraph X.B.3. 
The requirements in paragraph X.B.3 for submitting reports will vary 
according to certain time periods during a facility's lifetime. If a 
potential applicant for a COL that references this rule decides to 
depart from the generic DCD prior to submission of the application, 
then paragraph X.B.3.a will require that the updated DCD be submitted 
as part of the initial application for a license. Under paragraph 
X.B.3.b, the applicant may submit any subsequent updates to its plant-
specific DCD along with its amendments to the application provided that 
the submittals are made at least once per year.
    Paragraph X.B.3.b also requires semi-annual submission of the 
reports required by paragraphs X.B.1 and X.B.2 throughout the period of 
application review and construction. The NRC will use the information 
in the reports to support planning for the NRC's inspection and 
oversight during this phase, when the licensee is conducting detailed 
design, procurement of components and equipment, construction, and 
preoperational testing. In addition, the NRC will use the information 
in making its finding on ITAAC under Sec.  52.103(g), as well as any 
finding on interim operation under Section 189.a(1)(B)(iii) of the 
Atomic Energy Act of 1954, as amended. Once a facility begins operation 
(for a COL under 10 CFR part 52, after the Commission has made a 
finding under Sec.  52.103(g)), the frequency of reporting will be 
governed by the requirements in paragraph X.B.3.c.

VI. Public Comment Analysis

    The NRC prepared a summary and analysis of public comments received 
on the 2021 proposed rule, as referenced in the ``Availability of 
Documents'' section. The NRC received eight comment submissions during 
the public comment period that ended on October 14, 2021, and one late-
filed comment submission on October 15, 2021, that the NRC was able to 
include in its consideration for this final rule. A comment submission 
is a communication or document submitted to the NRC by an individual or 
entity, with one or more individual comments addressing a subject or 
issue. Private citizens provided four comment submissions, nuclear 
industry organizations provided two comment submissions, science 
advocacy groups provided two comment submissions, and a labor union 
provided one comment submission. Of the nine comments, six were in 
favor of the design certification rule, one was opposed, and the other 
two comment submittals posed questions but stated no preference for the 
outcome of the rule. Six of the nine comment submissions contained 
questions on technical aspects of the design, corrections to the 
statement of considerations, and interpretation of requirements.
    The public comment submittals are available on the Federal 
rulemaking website under Docket ID NRC-2017-0029. NRC's response to the 
public comments, including a summary of how NRC revised the proposed 
rule in response to public input, can be found in the public comment 
analysis document.

VII. Section-by-Section Analysis

    The following paragraphs describe the specific changes in this 
final rule: Section 52.11, Information collection requirements: Office 
of Management and Budget (OMB) approval.
    In Sec.  52.11, this final rule adds new appendix G to 10 CFR part 
52 to the list of information collection requirements in paragraph (b) 
of this section.

Appendix G to Part 52--Design Certification Rule for the NuScale 
Standard Design

    This final rule adds appendix G to 10 CFR part 52 to incorporate 
the NuScale standard design into the NRC's regulations. Applicants 
intending to construct and operate a plant using NuScale may do so by 
referencing the design certification rule.

VIII. Regulatory Flexibility Certification

    Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC 
certifies that this rule does not have a significant economic impact on 
a substantial number of small entities. This final rule affects only 
the licensing and operation of nuclear power plants. The companies that 
own these plants do not fall within the scope of the definition of 
``small entities'' set forth in the Regulatory Flexibility Act or the 
size standards established by the NRC (Sec.  2.810).

IX. Regulatory Analysis

    The NRC has not prepared a regulatory analysis for this final rule. 
The NRC prepares regulatory analyses for rulemakings that establish 
generic regulatory requirements applicable to all licensees. Design 
certifications are not generic rulemakings in the sense that design 
certifications do not establish standards or requirements with which 
all licensees must comply. Rather, design certifications are NRC 
approvals of specific nuclear power plant designs by rulemaking, which 
then may be voluntarily referenced by applicants for combined licenses. 
Furthermore, design certification rules are requested by an applicant 
for a design certification, rather than the NRC. Preparation of a 
regulatory analysis in this circumstance would not be useful because 
the design to be certified is proposed by the applicant rather than the 
NRC. For these reasons, the NRC concludes that preparation of a 
regulatory analysis is neither required nor appropriate.

X. Backfitting and Issue Finality

    The NRC has determined that this final rule does not constitute a 
backfit as defined in the backfit rule (Sec.  50.109), and that it is 
not inconsistent with any applicable issue finality provision in 10 CFR 
part 52.
    This initial design certification rule does not constitute 
backfitting as defined in the backfit rule (Sec.  50.109) because there 
are no operating licenses under 10 CFR part 50 referencing this design 
certification final rule.
    This initial design certification rule is not inconsistent with any 
applicable issue finality provision in 10 CFR part 52 because it does 
not impose new or changed requirements on existing design certification 
rules in appendices A through F to 10 CFR part 52, and no combined 
licenses, construction permits, or manufacturing licenses issued by the 
NRC at this time reference this design certification final rule.
    For these reasons, neither a backfit analysis nor a discussion 
addressing the issue finality provisions in 10 CFR part 52 was prepared 
for this final rule.

XI. Plain Writing

    The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal 
agencies to write documents in a clear, concise, well-organized manner 
that also follows other best practices appropriate to the subject or 
field and the intended audience. The NRC has written this

[[Page 3302]]

document to be consistent with the Plain Writing Act as well as the 
Presidential Memorandum, ``Plain Language in Government Writing,'' 
published June 10, 1998 (63 FR 31883).

XII. Environmental Assessment and Finding of No Significant Impact

    The NRC conducted an environmental assessment and has determined 
under the National Environmental Policy Act of 1969, as amended (NEPA), 
and the NRC's regulations in subpart A of 10 CFR part 51, that this 
final rule, if adopted, would not be a major Federal action 
significantly affecting the quality of the human environment and, 
therefore, an environmental impact statement is not required. The NRC's 
generic determination in this regard is reflected in Sec.  51.32(b)(1). 
The Commission has determined in Sec.  51.32 that there is no 
significant environmental impact associated with the issuance of a 
standard design certification or a design certification amendment, as 
applicable.
    The NRC's generic determination in this regard, as discussed in the 
2007 final rule amending 10 CFR parts 51 and 52 (72 FR 49351; August 
28, 2007), is based upon consideration that a design certification rule 
does not authorize the siting, construction, or operation of a facility 
referencing any particular design; it only codifies the NuScale design 
in a rule. The NRC will evaluate the environmental impacts and issue an 
environmental impact statement as appropriate under NEPA as part of the 
application for the construction and operation of a facility 
referencing any particular design certification rule.
    Consistent with Sec. Sec.  51.30(d) and 51.32(b), the NRC has 
prepared an environmental assessment for the NuScale design addressing 
various design alternatives to prevent and mitigate severe accidents. 
The environmental assessment is based, in part, upon the NRC's review 
of NuScale Power's evaluation of various design alternatives to prevent 
and mitigate severe accidents in Revision 5 of the DCA Part 3, 
``Application Applicant's Environmental Report--Standard Design 
Certification.'' Based on a review of NuScale Power's evaluation, the 
NRC concludes that (1) NuScale Power identified a reasonably complete 
set of potential design alternatives to prevent and mitigate severe 
accidents for the NuScale design and (2) none of the potential design 
alternatives appropriate at the design certification stage are 
justified on the basis of cost-benefit considerations. These issues are 
considered resolved for the NuScale design.
    Based on its own independent evaluation, the NRC concluded that 
none of the possible candidate design alternatives appropriate at this 
design certification stage are potentially cost beneficial for NuScale 
for accident events. This independent evaluation was based on 
reasonable treatment of costs, benefits, and sensitivities. The NRC's 
conclusion is applicable for sites with site characteristics that fall 
within the site parameters of the representative site specified in the 
NuScale environmental report. The NRC concludes that NuScale Power has 
adequately identified areas appropriate at this design certification 
stage where risk potentially could be reduced in a cost beneficial 
manner and that NuScale Power has adequately assessed whether the 
implementation of the identified potential severe accident mitigation 
design alternatives (SAMDAs) or candidate design alternatives would be 
cost beneficial for the representative site. As noted in the 
environmental assessment, SAMDA candidates for multi-unit sites are 
evaluated in the context of multiple NuScale reactor buildings, each 
with up to 12 power modules at the same site. Site-specific SAMDAs, 
multi-unit aspects, procedural and training SAMDAs, and the design 
element details of the reactor building crane will need to be assessed 
when an application for a specific site is submitted to construct and 
operate a NuScale power plant.
    The determination of this environmental assessment is that there 
will be no significant offsite impact to the public from this action. 
The environmental assessment is available as indicated under Section 
XVIII of this document.

XIII. Paperwork Reduction Act

    This final rule contains new or amended collections of information 
subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et 
seq.). The collections of information were approved by the Office of 
Management and Budget, approval number 3150-0151.
    The burden to the public for the information collections is 
estimated to average 130 hours per response, including the time for 
reviewing instructions, searching existing data sources, gathering and 
maintaining the data needed, and completing and reviewing the 
information collection.
    The information collection is being conducted to fulfill the 
requirements of a future applicant that references the design 
certification to maintain records of changes to and departures from the 
generic DCD, which are to be reflected in the plant-specific DCD. This 
information will be used by the NRC to fulfill its responsibilities in 
the licensing of nuclear power plants. Responses to this collection of 
information are mandatory. Confidential and proprietary information 
submitted to the NRC is protected in accordance with NRC regulations at 
Sec. Sec.  9.17(a) and 2.39(b).
    You may submit comments on any aspect of the information 
collections, including suggestions for reducing the burden, by the 
following methods:
     Federal rulemaking website: Go to https://www.regulations.gov search for Docket ID NRC-2017-0029.
     Mail comments to: FOIA, Library, and Information 
Collections Branch, Office of the Chief Information Officer, Mail Stop: 
T6-A10M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 
or to the OMB reviewer at: OMB Office of Information and Regulatory 
Affairs (3150-0151), Attn: Desk Officer for the Nuclear Regulatory 
Commission, 725 17th Street NW, Washington, DC 20503; email: 
[email protected].
Public Protection Notification
    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a collection of information unless the document requesting 
or requiring the collection displays a currently valid OMB control 
number.

XIV. Congressional Review Act

    This final rule is a rule as defined in the Congressional Review 
Act (5 U.S.C. 801-808). However, the Office of Management and Budget 
has not found it to be a major rule as defined in the Congressional 
Review Act.

XV. Agreement State Compatibility

    Under the ``Agreement State Program Policy Statement'' approved by 
the Commission on October 2, 2017, and published in the Federal 
Register on October 18, 2017 (82 FR 48535), this rule is classified as 
compatibility ``NRC.'' Compatibility is not required for Category 
``NRC'' regulations. The NRC program elements in this category are 
those that relate directly to areas of regulation reserved to the NRC 
by the AEA or the provisions of title 10 of the Code of Federal 
Regulations, and although an Agreement State may not adopt program 
elements reserved to the NRC, it may wish to inform its licensees of 
certain requirements via a mechanism that is consistent with a 
particular State's administrative procedure laws, but does not confer 
regulatory authority on the State.

[[Page 3303]]

XVI. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, 
Public Law 104-113, requires that Federal agencies use technical 
standards that are developed or adopted by voluntary consensus 
standards bodies unless the use of such a standard is inconsistent with 
applicable law or otherwise impractical. In this final rule, the NRC 
certifies the NuScale standard design for use in nuclear power plant 
licensing under 10 CFR parts 50 or 52. Design certifications are not 
generic rulemakings establishing a generally applicable standard with 
which all 10 CFR parts 50 and 52 nuclear power plant licensees must 
comply. Design certifications are Commission approvals of specific 
nuclear power plant designs by rulemaking. Furthermore, design 
certifications are initiated by an applicant for rulemaking, rather 
than by the NRC. This action does not constitute the establishment of a 
standard that contains generally applicable requirements.

XVII. Availability of Documents

    The documents identified in the following table are available to 
interested persons through one or more of the following methods, as 
indicated.

         Documents Related to NuScale Design Certification Rule
------------------------------------------------------------------------
                                                ADAMS accession No./web
                   Document                      link/Federal Register
                                                        citation
------------------------------------------------------------------------
SECY-22-0062, ``Final Rule: NuScale Small      ML22004A002
 Modular Reactor Design Certification (RIN
 3150-AJ98; NRC-2017-0029),'' July 1, 2022.
SECY-21-0004, ``Proposed Rule: NuScale Small   ML19353A003
 Modular Reactor Design Certification (RIN
 3150-AJ98; NRC-2017-0029),'' January 14,
 2021.
Staff Requirements Memorandum for SECY-21-     ML21126A153
 0004, ``Proposed Rule: NuScale Small Modular
 Reactor Design Certification (RIN 3150-AJ98;
 NRC-2017-0029),'' May 6, 2021.
Annotated Comment Submissions on Proposed      ML22045A21
 Rule: NuScale Small Modular Reactor Design
 Certification (NRC-2017-0029; RIN 3150-
 AJ98), June 2022.
Final Rule Comment Response Document for       ML22216A015
 NuScale Small Modular Reactor Design
 Certification (public comment analysis
 document), July 2022.
NuScale Power, LLC, Submittal of the NuScale   ML20225A071
 Standard Plant Design Certification
 Application, Revision 5, July 2020.
NuScale Standard Design Certification          ML20224A512
 Application, Part 3, ``Applicant's
 Environmental Report--Standard Design
 Certification,'' Revision 5, July 2020.
NuScale Power, LLC, Submittal of the NuScale   ML20205L562
 Standard Plant Design Certification
 Application, Revision 4.1, June 19, 2020.
NuScale Power, LLC, Submittal of the NuScale   ML19241A431
 Standard Plant Design Certification
 Application, Part 2, Tier 2, Revision 3,
 August 2019.
NuScale Power, LLC, Submittal of the NuScale   ML18310A345
 Standard Plant Design Certification
 Application, Part 2, Tier 2, Revision 2,
 October 2018.
NuScale Power, LLC, Topical report TR-0915-    ML19112A172
 17565, Revision 3, Accident Source Term
 Methodology, April 21, 2019.
Proposed Rule for the NuScale Small Modular    86 FR 34999
 Reactor Design Certification, July 1, 2021.
Extension of Comment Period for the Proposed   86 FR 47251
 Rule, August 24, 2021.
Docketing Notice for the NuScale Power, LLC,   82 FR 15717
 Design Certification Application (DCA),
 March 30, 2017.
Notification of Receipt of the NuScale Power,  82 FR 11372
 LLC, Design Certification Application (DCA),
 February 22, 2017.
NuScale Power, LLC, Submittal of the NuScale   ML17013A229
 Standard Plant Design Certification
 Application (NRC Project No. 0769), Revision
 0, December 2016.
NuScale Power, LLC, Submittal of NuScale       ML15258A846
 Preliminary Concept of Operations Summary
 and Response to NRC Questions on Control
 Room Activities, September 15, 2015.
Information on Differing Professional Opinion  ML22122A116
 (DPO) 2020-004, May 13, 2022.
------------------------------------------------------------------------
         Final Safety Evaluation Report and Supporting Documents
------------------------------------------------------------------------
NuScale DCA Final Safety Evaluation Report,    ML20023A318
 August 2020.
NRC Safety Evaluation for NuScale Power, LLC,  ML20044E199
 Topical Report, TR-0516-49422, ``Loss-of-
 Coolant,'' Revision 1, November 2019.
NRC Safety Evaluation for NuScale Power, LLC,  ML17340A524
 Topical Report, TR-0815-16497, Revision 1,
 ``Safety Classification of Passive Nuclear
 Power Plant Electrical Systems,'' December
 13, 2017.
NRC Safety Evaluation for NuScale Power, LLC,  ML19297G520
 Topical Report, TR-0915-17565, Rev. 3,
 ``Accident Source Term Methodology,''
 October 24, 2019.
NRR Response to Taskings in EDO DPO Appeal     ML22062A007
 Decision Concerning DPO-2020-004, May 13,
 2022.
------------------------------------------------------------------------
                          Environmental Reviews
------------------------------------------------------------------------
Final Environmental Assessment by the U.S.     ML22216A014
 Nuclear Regulatory Commission Relating to
 the Certification of the NuScale Standard
 Design, July 2022.
Environmental Assessment by the U.S. Nuclear   ML19303C179
 Regulatory Commission Relating to the
 Certification of the NuScale Standard
 Design, January 14, 2021.
Staff Technical Analysis in Support of the     ML19302E819
 NuScale Design Certification Environmental
 Assessment, August 4, 2020.
------------------------------------------------------------------------

[[Page 3304]]

 
  Commission Papers, Staff Requirement Memoranda, and Other Supporting
                                Documents
------------------------------------------------------------------------
SECY-11-0098, ``Operator Staffing for Small    ML111870574
 or Multi-Module Nuclear Power Plant
 Facilities,'' July 22, 2011.
SECY-17-0075, ``Planned Improvements in        ML16196A321
 Design Certification Tiered Information
 Designations,'' dated July 24, 2017.
SECY-18-0099, ``NuScale Power Exemption        ML18065A431
 Request from 10 CFR Part 50, Appendix A,
 General Design Criterion 27, `Combined
 Reactivity Control Systems Capability,' ''
 dated October 9, 2018.
SECY-19-0079, ``Staff Approach to Evaluate     ML19107A455
 Accident Source Terms for the NuScale Power
 Design Certification Application,'' August
 16, 2019.
SECY-77-439, ``Single Failure Criterion,''     ML060260236
 August 17, 1977.
SECY-93-087, ``Policy, Technical, and          ML003708021
 Licensing Issues Pertaining to Evolutionary
 and Advanced Light-Water Reactor (ALWR)
 Designs,'' April 2, 1993.
SRM-SECY-19-0036, ``Staff Requirements--SECY-  ML19183A408
 19-0036--Application of the Single Failure
 Criterion to NuScale Power LLC's Inadvertent
 Actuation Block Valves,'' July 2, 2019.
SRM-SECY-94-084, ``Policy and Technical        ML003708098
 Issues associated with the Regulatory
 Treatment of Non-Safety Systems and
 Implementation of Design Certification and
 Light-Water Reactor Design Issues,'' June
 30, 1994.
SRM-SECY-90-377, ``Requirements for Design     ML003707892
 Certification under 10 CFR part 52,''
 February 15, 1991.
Response to NuScale Power, LLC Key Issue       ML16229A522
 Resolution Letter, Supplemental Response
 Regarding Multi-Module Questions, October
 25, 2016.
Advisory Committee on Reactor Safeguards       ML20211M386
 (ACRS) Letter, ``Report on the Safety
 Aspects of the NuScale Small Modular
 Reactor,'' July 29, 2020.
American Society of Mechanical Engineers       https://webstore.ansi.org/
 Standard QME-1-2007, ``Qualification of        standards/asme/
 Active Mechanical Equipment Used in Nuclear    ansiasmeqme2007
 Power Plants,'' 2007.
NRC Regulatory Guide 1.100, Rev. 3, ``Seismic  ML091320468
 Qualification of Electrical and Active
 Mechanical Equipment and Functional
 Qualification of Active Mechanical Equipment
 for Nuclear Power Plants,'' September 2009.
NRC Regulatory Guide 1.206, Rev. 1,            ML18131A181
 ``Applications for Nuclear Power Plants,''
 October 2018.
NRC Agreement State Program Policy Statement,  82 FR 48535
 October 18, 2017.
Final Rule for Licenses, Certifications, and   72 FR 49351
 Approvals for Nuclear Power Plants (10 CFR
 parts 51 and 52), August 28, 2007.
Office of the Federal Register (OFR) Final     79 FR 66267
 Rule for Incorporation by Reference,
 November 7, 2014.
Presidential Memorandum, ``Plain Language in   63 FR 31883
 Government Writing,'' June 10, 1998.
Regulatory History of Design Certification,    ML003761550
 April 2000 \2\.
------------------------------------------------------------------------
                  NuScale Technical and Topical Reports
------------------------------------------------------------------------
ES-0304-1381-NP, Human-System Interface Style  ML19338E948
 Guide, Rev. 4, December 2019.
RP-0215-10815-NP, Concept of Operations, Rev.  ML19133A293
 3, May 2019.
RP-0316-17614-NP, Human Factors Engineering    ML16364A342
 Operating Experience Review Results Summary
 Report, Rev. 0, December 2016 \3\.
RP-0316-17615-NP, Human Factors Engineering    ML16364A342
 Functional Requirements Analysis and
 Function Allocation Results Summary Report,
 Rev. 0, December 2016 \3\.
RP-0316-17616-NP, Human Factors Engineering    ML19119A393
 Task Analysis Results Summary Report, Rev.
 2, April 2019.
RP-0316-17617-NP, Human Factors Engineering    ML17004A222
 Staffing and Qualifications Results Summary
 Report, Rev. 0, December 2016 \3\.
RP-0316-17618-NP, Human Factors Engineering    ML17004A222
 Treatment of Important Human Actions Results
 Summary Report, Rev. 0, December 2016 \3\.
RP-0316-17619-NP, Human Factors Engineering    ML19119A398
 Human-System Interface Design Results
 Summary Report, Rev. 2, April 2019.
RP-0516-49116-NP, Control Room Staffing Plan   ML16364A356
 Validation Results, Rev. 1, December 2016.
RP-0914-8534-NP, Human Factors Engineering     ML19119A342
 Program Management Plan, Rev. 5, April 2019.
RP-0914-8543-NP, Human Factors Verification    ML19119A372
 and Validation Implementation Plan, Rev. 5,
 April 2019.
RP-0914-8544-NP, Human Factors Engineering     ML19331A910
 Design Implementation Plan, Rev. 4, November
 2019.
RP-1018-61289-NP, Human Factors Engineering    ML19212A773
 Verification and Validation Results Summary
 Report, Rev. 1, July 2019.
RP-1215-20253-NP, Control Room Staffing Plan   ML16364A353
 Validation Methodology, Rev. 3, December
 2016.
TR-0116-20781-NP, Fluence Calculation          ML19183A485
 Methodology and Results, Rev. 1, July 2019.
TR-0116-20825-NP-A, Applicability of AREVA     ML18040B306
 Fuel Methodology for the NuScale Design,
 Rev. 1, June 2016.
TR-0116-21012-NP-A, NuScale Power Critical     ML18360A632
 Heat Flux Correlations, Rev. 1, December
 2018.
TR-0316-22048-NP, Nuclear Steam Supply System  ML20141M764
 Advanced Sensor Technical Report, Rev. 3,
 May 2020.
TR-0515-13952-NP-A, Risk Significance          ML16284A016
 Determination, Rev. 0, October 2016.
TR-0516-49084-NP, Containment Response         ML20141L808
 Analysis Methodology Technical Report, Rev.
 3, May 2020.

[[Page 3305]]

 
TR-0516-49416-NP-A, Non-Loss-of-Coolant        ML20191A281
 Accident Analysis Methodology, Rev. 3, July
 2020.
TR-0516-49417-NP-A, Evaluation Methodology     ML20078Q094
 for Stability Analysis of the NuScale Power
 Module, Rev. 1, March 2020.
TR-0516-49422-NP-A, Loss-of-Coolant Accident   ML20189A644
 Evaluation Model, Rev. 2, July 2020.
TR-0616-48793-NP-A, Nuclear Analysis Codes     ML18348B036
 and Methods Qualification, Rev. 1, November
 2018.
TR-0616-49121-NP, NuScale Instrument Setpoint  ML20141M114
 Methodology Technical Report, Rev. 3, May
 2020.
TR-0716-50350-NP-A, Rod Ejection Accident      ML20168B203
 Methodology, Rev. 1, June 2020.
TR-0716-50351-NP-A, NuScale Applicability of   ML20122A248
 AREVA Method for the Evaluation of Fuel
 Assembly Structural Response to Externally
 Applied Forces, Rev. 1, April 2020.
TR-0716-50424-NP, Combustible Gas Control,     ML19091A232
 Rev. 1, March 2019.
TR-0716-50439-NP, NuScale Comprehensive        ML19212A776
 Vibration Assessment Program Analysis
 Technical Report, Rev. 2, July 2019.
TR-0815-16497-NP-A, Safety Classification of   ML18054B607
 Passive Nuclear Power Plant Electrical
 Systems Topical Report, Rev. 1, January 2018.
TR-0816-49833-NP, Fuel Storage Rack Analysis,  ML18310A154
 Rev. 1, November 2018.
TR-0816-50796-NP, Loss of Large Areas Due to   ML19165A294
 Explosions and Fires Assessment, Rev. 1,
 June 2019.
TR-0816-50797 (NuScale Nonproprietary),        ML19302H598
 Mitigation Strategies for Loss of All AC
 Power Event, Rev. 3, October 2019.
TR-0816-51127-NP, NuFuel-HTP2TM Fuel and       ML19353A719
 Control Rod Assembly Designs, Rev. 3,
 December 2019.
TR-0818-61384-NP, Pipe Rupture Hazards         ML19212A682
 Analysis, Rev. 2, July 2019.
TR-0915-17564-NP-A, Subchannel Analysis        ML19067A256
 Methodology, Rev. 2, February 2019.
TR-0915-17565-NP-A, Accident Source Term       ML20057G132
 Methodology, Rev. 4, February 2020.
TR-0916-51299-NP, Long-Term Cooling            ML20141L816
 Methodology, Rev. 3, May 2020.
TR-0916-51502-NP, NuScale Power Module         ML19093B850
 Seismic Analysis, Rev. 2, April 2019.
TR-0917-56119-NP, CNV Ultimate Pressure        ML19158A382
 Integrity, Rev. 1, June 2019.
TR-0918-60894-NP, Comprehensive Vibration      ML19214A248
 Assessment Program Measurement and
 Inspection Plan Technical Report, Rev. 1,
 August 2019.
TR-1010-859-NP-A, NuScale Topical Report:      ML20176A494
 Quality Assurance Program Description for
 the NuScale Power Plant, Rev. 5, May 2020.
TR-1015-18177-NP, Pressure and Temperature     ML18298A304
 Limits Methodology, Rev. 2, October 2018.
TR-1015-18653-NP-A, Design of the Highly       ML17256A892
 Integrated Protection System Platform
 Topical Report, Rev. 2, May 2017.
TR-1016-51669-NP, NuScale Power Module Short-  ML19211D411
 Term Transient Analysis, Rev. 1, July 2019.
TR-1116-51962-NP, NuScale Containment Leakage  ML19149A298
 Integrity Assurance, Rev. 1, May 2019.
TR-1116-52065-NP, Effluent Release (GALE       ML18317A364
 Replacement) Methodology and Results, Rev.
 1, November 2018.
------------------------------------------------------------------------

    The NRC may post materials related to this document, including 
public comments, on the Federal rulemaking website at https://www.regulations.gov under Docket ID NRC-2017-0029. In addition, the 
Federal rulemaking website allows members of the public to receive 
alerts when changes or additions occur in a docket folder. To 
subscribe: (1) navigate to the docket folder (NRC-2017-0029); (2) click 
the ``Subscribe'' link; and (3) enter an email address and click on the 
``Subscribe'' link.
---------------------------------------------------------------------------

    \2\ The regulatory history of the NRC's design certification 
reviews is a package of documents that is available in the NRC's PDR 
and NRC Library. This history spans the period during which the NRC 
simultaneously developed the regulatory standards for reviewing 
these designs and the form and content of the rules that certified 
the designs.
    \3\ The duplicate ADAMS Accession Nos. ML16364A342 and 
ML17004A222 are intentional and indicate when multiple reports are 
part of a single submittal.
---------------------------------------------------------------------------

XVIII. Incorporation by Reference--Reasonable Availability to 
Interested Parties

    The NRC is incorporating by reference the NuScale DCA, Revision 5. 
As described in the ``Discussion'' sections of this document, the 
generic DCD includes Tier 1 and Tier 2 information (including the 
technical and topical reports referenced in Chapter 1) and generic 
technical specifications in order to effectively control this 
information and facilitate its incorporation by reference into the 
rule. NuScale Power submitted Revision 5 of the DCA to the NRC in July 
2020.
    The NRC is required by law to obtain approval for incorporation by 
reference from the Office of the Federal Register (OFR). The OFR's 
requirements for incorporation by reference are set forth in 1 CFR part 
51. On November 7, 2014, the OFR adopted changes to its regulations 
governing incorporation by reference (79 FR 66267). The OFR regulations 
require an agency to discuss, in the preamble of the final rule, the 
ways that the materials it incorporates by reference are reasonably 
available to interested parties and how interested parties can obtain 
the materials. The discussion in this section complies with the 
requirement for final rules as set forth in 1 CFR 51.5(a)(1).
    The NRC considers ``interested parties'' to include all potential 
NRC stakeholders, not only the individuals and entities regulated or 
otherwise subject to the NRC's regulatory oversight. These NRC 
stakeholders are not a homogenous group but vary with respect to the 
considerations for determining reasonable availability. Therefore, the 
NRC distinguishes between different classes of interested parties for 
the purposes of determining whether the material is ``reasonably 
available.'' The NRC considers the following to be classes of 
interested parties in NRC rulemakings with regard

[[Page 3306]]

to the material to be incorporated by reference:
     Individuals and small entities regulated or otherwise 
subject to the NRC's regulatory oversight (this class also includes 
applicants and potential applicants or licenses and other NRC 
regulatory approvals) and who are subject to the material to be 
incorporated by reference by rulemaking. In this context, ``small 
entities'' has the same meaning as a ``small entity'' under Sec.  
2.810.
     Large entities otherwise subject to the NRC's regulatory 
oversight (this class also includes applicants and potential applicants 
for licenses and other NRC regulatory approvals) and who are subject to 
the material to be incorporated by reference by rulemaking. In this 
context, ``large entities'' are those which do not qualify as a ``small 
entity'' under Sec.  2.810.
     Non-governmental organizations with institutional 
interests in the matters regulated by the NRC.
     Other Federal agencies, States, and local governmental 
bodies (within the meaning of Sec.  2.315(c)).
     Federally-recognized and State-recognized \4\ Indian 
tribes.
---------------------------------------------------------------------------

    \4\ State-recognized Indian tribes are not within the scope of 
10 CFR 2.315(c). However, for purposes of the NRC's compliance with 
1 CFR 51.5, ``interested parties'' includes a broad set of 
stakeholders, including State-recognized Indian tribes.
---------------------------------------------------------------------------

     Members of the general public (i.e., individual, 
unaffiliated members of the public who are not regulated or otherwise 
subject to the NRC's regulatory oversight) who may wish to gain access 
to the materials which the NRC incorporates by reference by rulemaking 
in order to participate in the rulemaking process.
    The NRC makes the materials incorporated by reference available for 
inspection to all interested parties, by appointment, at the NRC 
Technical Library, which is located at Two White Flint North, 11545 
Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-7000; 
email: [email protected]. In addition, as described in Section 
XVIII of this document, documents related to this final rule are 
available online in the NRC's ADAMS Public Documents collection at 
https://www.nrc.gov/reading-rm/adams.html.
    The NRC concludes that the materials the NRC is incorporating by 
reference in this final rule are reasonably available to all interested 
parties because the materials are available in multiple ways and in a 
manner consistent with their interest in the materials.

List of Subjects in 10 CFR Part 52

    Administrative practice and procedure, Antitrust, Combined license, 
Early site permit, Emergency planning, Fees, Incorporation by 
reference, Inspection, Issue finality, Limited work authorization, 
Nuclear power plants and reactors, Probabilistic risk assessment, 
Prototype, Reactor siting criteria, Redress of site, Penalties, 
Reporting and recordkeeping requirements, Standard design, Standard 
design certification.

    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; the Nuclear Waste Policy Act of 1982, as 
amended; and 5 U.S.C. 552 and 553, the NRC is amending 10 CFR part 52 
as follows:

PART 52--LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER 
PLANTS

0
1. The authority citation for part 52 continues to read as follows:

    Authority: Atomic Energy Act of 1954, secs. 103, 104, 147, 149, 
161, 181, 182, 183, 185, 186, 189, 223, 234 (42 U.S.C. 2133, 2134, 
2167, 2169, 2201, 2231, 2232, 2233, 2235, 2236, 2239, 2273, 2282); 
Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 
U.S.C. 5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.


Sec.  52.11   [Amended]

0
2. In Sec.  52.11(b), remove the phrase ``appendices A, B, C, D, E, F, 
and N of this part'' and add, in its place, the phrase ``appendices A, 
B, C, D, E, F, G, and N of this part''.

0
3. Add appendix G to part 52 to read as follows:

Appendix G to Part 52--Design Certification Rule for NuScale

I. Introduction

    Appendix G constitutes the standard design certification for the 
NuScale design (hereinafter referred to as NuScale), in accordance 
with 10 CFR part 52, subpart B. The applicant for this standard 
design certification NuScale is NuScale Power, LLC.

II. Definitions

    A. Generic design control document (generic DCD) means the 
documents containing the Tier 1 and Tier 2 information (including 
the technical and topical reports referenced in Chapter 1) and 
generic technical specifications that are incorporated by reference 
into this appendix.
    B. Generic technical specifications (generic TS) means the 
information required by 10 CFR 50.36 and 50.36a for the portion of 
the plant that is within the scope of this appendix.
    C. Plant-specific DCD means that portion of the combined license 
(COL) final safety analysis report (FSAR) that sets forth both the 
generic DCD information and any plant-specific changes to generic 
DCD information.
    D. Tier 1 means the portion of the design-related information 
contained in the generic DCD that is approved and certified by this 
appendix (Tier 1 information). The design descriptions, interface 
requirements, and site parameters are derived from Tier 2 
information. Tier 1 information includes:
    1. Definitions and general provisions;
    2. Design descriptions;
    3. Inspections, tests, analyses, and acceptance criteria 
(ITAAC);
    4. Significant site parameters; and
    5. Significant interface requirements.
    E. Tier 2 means the portion of the design-related information 
contained in the generic DCD that is approved but not certified by 
this appendix (Tier 2 information). Compliance with Tier 2 is 
required, but generic changes to and plant-specific departures from 
Tier 2 are governed by Section VIII of this appendix. Compliance 
with Tier 2 provides a sufficient, but not the only acceptable, 
method for complying with Tier 1. Compliance methods differing from 
Tier 2 must satisfy the change process in Section VIII of this 
appendix. Regardless of these differences, an applicant or licensee 
must meet the requirement in paragraph III.B of this appendix to 
reference Tier 2 when referencing Tier 1. Tier 2 information 
includes:
    1. Information required by Sec.  52.47(a) and (c), with the 
exception of generic TS and conceptual design information;
    2. Supporting information on the inspections, tests, and 
analyses that will be performed to demonstrate that the acceptance 
criteria in the ITAAC have been met; and
    3. COL action items (COL license information) identify certain 
matters that must be addressed in the site-specific portion of the 
FSAR by an applicant who references this appendix. These items 
constitute information requirements but are not the only acceptable 
set of information in the FSAR. An applicant may depart from or omit 
these items, provided that the departure or omission is identified 
and justified in the FSAR. After issuance of a construction permit 
or COL, these items are not requirements for the licensee unless 
such items are restated in the FSAR.
    F. Departure from a method of evaluation described in the plant-
specific DCD used in establishing the design bases or in the safety 
analyses means:
    1. Changing any of the elements of the method described in the 
plant-specific DCD unless the results of the analysis are 
conservative or essentially the same; or
    2. Changing from a method described in the plant-specific DCD to 
another method unless that method has been approved by the NRC for 
the intended application.
    G. Nuclear power unit, as applied to this certified design, 
means a nuclear power module and associated equipment necessary for 
electric power generation and includes those structures, systems, 
and components required to provide reasonable assurance the facility 
can be operated without undue risk to the health and safety of the 
public.
    H. All other terms in this appendix have the meaning set out in 
10 CFR 50.2, 10 CFR

[[Page 3307]]

52.1, or Section 11 of the Atomic Energy Act of 1954, as amended, as 
applicable.

III. Scope and Contents

    A. Incorporation by reference.
    1. Certain material listed in paragraph III.A.2 of this appendix 
is incorporated by reference into this appendix G with the approval 
of the Director of the Federal Register in accordance with 5 U.S.C. 
552(a) and 1 CFR part 51. All approved incorporation by reference 
(IBR) material in paragraph III.A.2 of this appendix may be obtained 
from NuScale Power, LLC, 6650 SW Redwood Lane, Suite 210, Portland, 
Oregon 97224, telephone: 1-971-371-1592, email: 
[email protected], and can be inspected as follows:
    a. Contact the U.S. Nuclear Regulatory Commission at: U.S. 
Nuclear Regulatory Commission, Two White Flint North, 11545 
Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-7000; 
email: [email protected]; https://www.nrc.gov/reading-rm/pdr.html.
    b. Access ADAMS and view the material online in the NRC Library 
at https://www.nrc.gov/reading-rm/adams.html. In ADAMS, search under 
ADAMS Accession No. ML20225A071. The material is available in the 
ADAMS Public Documents collection.
    c. If you do not have access to ADAMS or if you have problems 
accessing documents located in ADAMS, contact the NRC's Public 
Document Room (PDR) reference staff at 1-800-397-4209, 301-415-3747, 
or by email at [email protected].
    d. For information on the availability of this material at the 
National Archives and Records Administration, visit 
www.archives.gov/federal-register/cfr/ibr-locations.html or email: 
[email protected].
    2. Material incorporated by reference.
    a. NuScale Standard Plant Design Certification Application, 
Certified Design Descriptions and Inspections, Tests, Analyses, & 
Acceptance Criteria (ITAAC), Part 2--Tier 1, Revision 5, July 2020.
    b. NuScale Standard Plant Design Certification Application, Part 
2--Tier 2, Revision 5, July 2020, including:
    i. Chapter One, Introduction and General Description of the 
Plant.
    ii. Chapter Two, Site Characteristics and Site Parameters.
    iii. Chapter Three, Design of Structures, Systems, Components 
and Equipment.
    iv. Chapter Four, Reactor.
    v. Chapter Five, Reactor Coolant System and Connecting Systems.
    vi. Chapter Six, Engineered Safety Features.
    vii. Chapter Seven, Instrumentation and Controls.
    viii. Chapter Eight, Electric Power.
    ix. Chapter Nine, Auxiliary Systems.
    x. Chapter Ten, Steam and Power Conversion System.
    xi. Chapter Eleven, Radioactive Waste Management.
    xii. Chapter Twelve, Radiation Protection.
    xiii. Chapter Thirteen, Conduct of Operations.
    xiv. Chapter Fourteen, Initial Test Program and Inspections, 
Tests, Analyses, and Acceptance Criteria.
    xv. Chapter Fifteen, Transient and Accident Analyses.
    xvi. Chapter Sixteen, Technical Specifications.
    xvii. Chapter Seventeen, Quality Assurance and Reliability 
Assurance.
    xviii. Chapter Eighteen, Human Factors Engineering.
    xix. Chapter Nineteen, Probabilistic Risk Assessment and Severe 
Accident Evaluation.
    xx. Chapter Twenty, Mitigation of Beyond-Design-Basis Events.
    xxi. Chapter Twenty-One, Multi-Module Design Considerations.
    c. DCA Part 4, Volume 1, Revision 5.0, Generic Technical 
Specifications, NuScale Nuclear Power Plants, Volume 1: 
Specifications.
    d. DCA Part 4, Volume 2, Revision 5.0, Generic Technical 
Specifications, NuScale Nuclear Power Plants, Volume 2: Bases.
    e. ES-0304-1381-NP, Human-System Interface Style Guide, December 
2019, Revision 4.
    f. RP-0215-10815-NP, Concept of Operations, May 2019, Revision 
3.
    g. RP-0316-17614-NP, Human Factors Engineering Operating 
Experience Review Results Summary Report, December 7, 2016, Revision 
0.
    h. RP-0316-17615-NP, Human Factors Engineering Functional 
Requirements Analysis and Function Allocation Results Summary 
Report, December 2, 2016, Revision 0.
    i. RP-0316-17616-NP, Human Factors Engineering Task Analysis 
Results Summary Report, April 2019, Revision 2.
    j. RP-0316-17617-NP, Human Factors Engineering Staffing and 
Qualifications Results Summary Report, December 2, 2016, Revision 0.
    k. RP-0316-17618-NP, Human Factors Engineering Treatment of 
Important Human Actions Results Summary Report, December 2, 2016, 
Revision 0.
    l. RP-0316-17619-NP, Human Factors Engineering Human-System 
Interface Design Results Summary Report, April 2019, Revision 2.
    m. RP-0516-49116-NP, Control Room Staffing Plan Validation 
Results, December 2, 2016, Revision 1.
    n. RP-0914-8534-NP, Human Factors Engineering Program Management 
Plan, April 2019, Revision 5.
    o. RP-0914-8543-NP, Human Factors Verification and Validation 
Implementation Plan, April 2019, Revision 5.
    p. RP-0914-8544-NP, Human Factors Engineering Design 
Implementation Plan, November 2019, Revision 4.
    q. RP-1018-61289-NP, Human Factors Engineering Verification and 
Validation Results Summary Report, July 2019, Revision 1.
    r. RP-1215-20253-NP, Control Room Staffing Plan Validation 
Methodology, December 2, 2016, Revision 3.
    s. TR-0116-20781-NP, Fluence Calculation Methodology and 
Results, July 2019, Revision 1.
    t. TR-0116-20825-NP-A, Applicability of AREVA Fuel Methodology 
for the NuScale Design, June 2016, Revision 1.
    u. TR-0116-21012-NP-A, NuScale Power Critical Heat Flux 
Correlations, December 2018, Revision 1.
    v. TR-0316-22048-NP, Nuclear Steam Supply System Advanced Sensor 
Technical Report, May 2020, Revision 3.
    w. TR-0515-13952-NP-A, Risk Significance Determination, October 
2016, Revision 0.
    x. TR-0516-49084-NP, Containment Response Analysis Methodology 
Technical Report, May 2020, Revision 3.
    y. TR-0516-49416-NP-A, Non-Loss-of-Coolant Accident Analysis 
Methodology, July 2020, Revision 3.
    z. TR-0516-49417-NP-A, Evaluation Methodology for Stability 
Analysis of the NuScale Power Module, March 2020, Revision 1.
    aa. TR-0516-49422-NP-A, Loss-of-Coolant Accident Evaluation 
Model, July 2020, Revision 2.
    ab. TR-0616-48793-NP-A, Nuclear Analysis Codes and Methods 
Qualification, November 2018, Revision 1.
    ac. TR-0616-49121-NP, NuScale Instrument Setpoint Methodology 
Technical Report, May 2020, Revision 3.
    ad. TR-0716-50350-NP-A, Rod Ejection Accident Methodology, June 
2020, Revision 1.
    ae. TR-0716-50351-NP-A, NuScale Applicability of AREVA Method 
for the Evaluation of Fuel Assembly Structural Response to 
Externally Applied Forces, April 2020, Revision 1.
    af. TR-0716-50424-NP, Combustible Gas Control, March 2019, 
Revision 1.
    ag. TR-0716-50439-NP, NuScale Comprehensive Vibration Assessment 
Program Analysis Technical Report, July 2019, Revision 2.
    ah. TR-0815-16497-NP-A, Safety Classification of Passive Nuclear 
Power Plant Electrical Systems, January 2018, Revision 1.
    ai. TR-0816-49833-NP, Fuel Storage Rack Analysis, November 2018, 
Revision 1.
    aj. TR-0816-50796-NP, Loss of Large Areas Due to Explosions and 
Fires Assessment, June 2019, Revision 1.
    ak. TR-0816-50797, Mitigation Strategies for Loss of All AC 
Power Event [NuScale Nonproprietary], October 2019, Revision 3.
    al. TR-0816-51127-NP, NuFuel-HTP2TM Fuel and Control 
Rod Assembly Designs, December 2019, Revision 3.
    am. TR-0818-61384-NP, Pipe Rupture Hazards Analysis, July 2019, 
Revision 2.
    an. TR-0915-17564-NP-A, Subchannel Analysis Methodology, 
February 2019, Revision 2.
    ao. TR-0915-17565-NP-A, Accident Source Term Methodology, 
February 2020, Revision 4.
    ap. TR-0916-51299-NP, Long-Term Cooling Methodology, May 2020, 
Revision 3.
    aq. TR-0916-51502-NP, NuScale Power Module Seismic Analysis, 
April 2019, Revision 2.
    ar. TR-0917-56119-NP, CNV Ultimate Pressure Integrity, June 
2019, Revision 1.
    as. TR-0918-60894-NP, NuScale Comprehensive Vibration Assessment 
Program Measurement and Inspection Plan Technical Report, August 
2019, Revision 1.
    at. NP-TR-1010-859-NP-A, NuScale Topical Report: Quality 
Assurance Program

[[Page 3308]]

Description for the NuScale Power Plant, May 2020, Revision 5.
    au. TR-1015-18177-NP, Pressure and Temperature Limits 
Methodology, October 2018, Revision 2.
    av. TR-1015-18653-NP-A, Design of the Highly Integrated 
Protection System Platform, May 2017, Revision 2.
    aw. TR-1016-51669-NP, NuScale Power Module Short-Term Transient 
Analysis, July 2019, Revision 1.
    ax. TR-1116-51962-NP, NuScale Containment Leakage Integrity 
Assurance, May 2019, Revision 1.
    ay. TR-1116-52065-NP, Effluent Release (GALE Replacement) 
Methodology and Results, November 2018, Revision 1.
    B.1. An applicant or licensee referencing this appendix, in 
accordance with Section IV of this appendix, shall incorporate by 
reference and comply with the requirements of this appendix except 
as otherwise provided in this appendix.
    2. Conceptual design information, as set forth in the design 
certification application Part 2, Tier 2, Section 1.2, and the 
discussion of ``first principles'' contained in design certification 
application Part 2, Tier 2, Section 14.3.2, are not incorporated by 
reference into this appendix.
    C. If there is a conflict between Tier 1 and Tier 2 of the DCD, 
then Tier 1 controls.
    D. If there is a conflict between the generic DCD and either the 
application for the design certification of NuScale or the final 
safety evaluation report related to certification of the NuScale 
standard design, then the generic DCD controls.
    E. Design activities for structures, systems, and components 
that are wholly outside the scope of this appendix may be performed 
using site characteristics, provided the design activities do not 
affect the DCD or conflict with the interface requirements.

IV. Additional Requirements and Restrictions

    A. An applicant for a COL that wishes to reference this appendix 
shall, in addition to complying with the requirements of Sec. Sec.  
52.77, 52.79, and 52.80, comply with the following requirements:
    1. Incorporate by reference, as part of its application, this 
appendix.
    2. Include, as part of its application:
    a. A plant-specific DCD containing the same type of information 
and using the same organization and numbering as the generic DCD for 
NuScale, either by including or incorporating by reference the 
generic DCD information, and as modified and supplemented by the 
applicant's exemptions and departures;
    b. The reports on departures from and updates to the plant-
specific DCD required by paragraph X.B of this appendix;
    c. Plant-specific TS, consisting of the generic and site-
specific TS that are required by 10 CFR 50.36 and 50.36a;
    d. Information demonstrating that the site characteristics fall 
within the site parameters and that the interface requirements have 
been met;
    e. Information that addresses the COL action items;
    f. Information required by Sec.  52.47(a) that is not within the 
scope of this appendix;
    g. Information demonstrating that necessary shielding to limit 
radiological dose consistent with the radiation zones specified in 
design certification application Part 2, Tier 2, Chapter 12, Figure 
12.3-1, ``Reactor Building Radiation Zone Map,'' is provided to 
account for penetrations in the radiation shield wall between the 
power module bay and the reactor building steam gallery area;
    h. Information demonstrating that the requirements of 10 CFR 
50.34(f)(2)(xxviii) are met with respect to potential radiological 
releases under accident conditions from the systems used for post-
accident hydrogen and oxygen monitoring described in design 
certification application Part 2, Tier 2, Section 6.2.5; information 
demonstrating that post-accident leakage from these systems does not 
result in the total main control room dose exceeding the dose 
criteria for the surrogate event with significant core damage, which 
may include use of design features compliant with 10 CFR 
50.34(f)(2)(vii), as appropriate; and information demonstrating that 
post-accident leakage from these systems does not result in the 
total dose for the surrogate event with significant core damage 
exceeding the offsite dose criteria, as required by 10 CFR 
52.47(a)(2)(iv); and
    i. Information demonstrating that the requirements of 10 CFR 
52.47(a)(2)(iv) and General Design Criterion (GDC) 4 and GDC 31 of 
appendix A to 10 CFR part 50 are met with respect to the structural 
and leakage integrity of the steam generator tubes that might be 
compromised by effects from density wave oscillations in the 
secondary fluid system, including the method of analysis to predict 
the thermal-hydraulic conditions of the steam generator secondary 
fluid system and resulting loads, stresses, and deformations from 
density wave oscillations and reverse flow. This information must be 
consistent with the other design information regarding steam 
generator integrity contained in design certification application 
Part 2, Tier 2, Sections 3.9.2 and 5.4.1.
    3. Include, in the plant-specific DCD, the sensitive, 
unclassified, non-safeguards information (including proprietary 
information and security-related information) and safeguards 
information referenced in the NuScale generic DCD.
    4. Include, as part of its application, a demonstration that an 
entity other than NuScale Power, LLC, is qualified to supply the 
NuScale generic DCD, unless NuScale Power, LLC, supplies the design 
for the applicant's use.
    B. The Commission reserves the right to determine in what manner 
this appendix may be referenced by an applicant for a construction 
permit or operating license under 10 CFR part 50.
    C. A licensee referencing the NuScale design certification is 
exempt from portions of the following regulation:
    1. Paragraph (m) of 10 CFR 50.54--Minimum Staffing. In lieu of 
these requirements, a licensee that references this appendix must 
comply with the following:
    a. A senior operator licensed pursuant to part 55 of this 
chapter shall be present at the facility or readily available on 
call at all times during its operation, and shall be present at the 
facility during initial startup and approach to power, recovery from 
an unplanned or unscheduled shutdown or significant reduction in 
power, and refueling, or as otherwise prescribed in the facility 
license.
    b. Licensees shall meet the following requirements:
    i. Each licensee shall meet the minimum licensed operator 
staffing requirements identified in Table 1:

 Table 1--Minimum Requirements per Shift for On-Site Staffing of NuScale
  Power Plants by Operators and Senior Operators Licensed Under 10 CFR
                                 Part 55
------------------------------------------------------------------------
Number of units operating (a                               One to twelve
    nuclear power unit is                                      units
 considered to be operating                              ---------------
 when it is in MODE 1, 2, or           Position
 3 as defined by the unit's                                 One control
  technical specifications)                                    room
------------------------------------------------------------------------
None........................  Senior operator...........               1
                              Operator..................               2
One to twelve...............  Senior operator...........               3
                              Operator..................               3
------------------------------------------------------------------------
Source: Design Certification Application, Part 7, Section 6.1.3,
  ``Requested Action.''

    ii. Each facility licensee shall have at its site a person 
holding a senior operator license for all fueled units at the site 
who is assigned responsibility for overall plant operation at all 
times there is fuel in any unit. At all times any module is fueled, 
regardless of mode, there must be a licensed operator or senior 
operator in the control room.
    iii. When a nuclear power unit is in MODE 1, 2, or 3, as defined 
by the unit's technical

[[Page 3309]]

specifications, each licensee shall have a person holding a senior 
operator license for the nuclear power unit in the control room at 
all times. In addition to this senior operator, a second person who 
is either a licensed operator or licensed senior operator shall be 
present at the controls at all times. A third person who is either a 
licensed operator or licensed senior operator shall be in the 
control room envelope at all times.
    iv. Each licensee shall have present, during alteration or 
movement of the core of a nuclear power unit (including fuel 
loading, fuel transfer, or movement of a module that contains fuel), 
a person holding a senior operator license or a senior operator 
license limited to fuel handling to directly supervise the activity 
and, during this time, the licensee shall not assign other duties to 
this person.
    2. Appendix J to 10 CFR part 50, Type A testing--Primary Reactor 
Containment Leakage Testing for Water-Cooled Power Reactors.

V. Applicable Regulations

    A. Except as indicated in paragraph B of this section, the 
regulations that apply to NuScale are in 10 CFR parts 20, 50, 52, 
73, and 100, codified as of February 21, 2023, that are applicable 
and technically relevant, as described in the final safety 
evaluation report.
    B. The NuScale design is exempt from portions of the following 
regulations:
    1. Paragraph (f)(2)(vi) of 10 CFR 50.34 and 10 CFR 50.46a--High 
point venting for the reactor coolant system and reactor pressure 
vessel head.
    2. Paragraph (f)(2)(viii) of 10 CFR 50.34--Post-accident 
sampling of the reactor coolant system and containment.
    3. Paragraph (f)(2)(xiii) of 10 CFR 50.34--Power supplies for 
pressurizer heaters.
    4. Paragraph (f)(2)(xiv)(E) of 10 CFR 50.34--Automatic closing 
of containment isolation systems on a high radiation signal.
    5. Paragraph (f)(2)(xx) of 10 CFR 50.34--Power from vital buses 
and emergency power sources for pressurizer level indication.
    6. Paragraph (c)(2) of 10 CFR 50.44--Combustible gas control.
    7. Paragraph (a)(1)(i) of 10 CFR 50.46--Applicability limited to 
reactor designs that use zircaloy or ZIRLO fuel rod cladding 
material.
    8. Paragraph (c)(1) of 10 CFR 50.62--Diverse equipment to 
initiate a turbine trip under conditions indicative of an 
anticipated transient without scram.
    9 Appendix A of 10 CFR part 50--Electric Power Systems GDCs:
    a. GDC 17--Electric power systems for safety-related functions;
    b. GDC 18--Design to permit periodic inspection and testing of 
electric power systems;
    c. GDC 34--Electric power systems for residual heat removal;
    d. GDC 35--Electric power systems for emergency core cooling;
    e. GDC 38--Electric power systems for containment heat removal;
    f. GDC 41--Electric power systems for containment atmosphere 
cleanup; and
    g. GDC 44--Electric power systems for cooling.
    10. Appendix A to 10 CFR part 50, GDC 19--Equipment outside the 
control room with capability for cold shutdown of the reactor.
    11. Appendix A to 10 CFR part 50, GDC 27--Demonstration of long-
term shutdown under post-accident conditions with an assumed worst 
rod stuck out.
    12. Appendix A to 10 CFR part 50, GDC 33--Reactor coolant makeup 
for protection against small breaks in the reactor coolant pressure 
boundary.
    13. Appendix A to 10 CFR part 50, GDC 40--Periodic pressure and 
functional testing of containment heat removal system.
    14. Appendix A to 10 CFR part 50, GDC 52--Design to allow 
periodic containment leakage rate testing.
    15. Appendix A of 10 CFR part 50, GDCs 55, 56, and 57--
Containment Isolation:
    a. GDC 55--Isolation valves for certain reactor coolant pressure 
boundary lines penetrating containment;
    b. GDC 56--Isolation valves for certain primary containment 
lines; and
    c. GDC 57--Isolation valves for certain closed systems lines.
    16. Appendix K to 10 CFR part 50--Emergency Core Cooling System 
Evaluation Models:
    a. Section I.A.4--Heat generation rates from radioactive decay 
of fission products;
    b. Section I.A.5--Rate of energy release, hydrogen generation, 
and cladding oxidation from the metal/water reaction;
    c. Section I.B--Predicting cladding swelling and rupture;
    d. Section I.C.1.b--Calculation of the discharge rate for all 
times after the discharging fluid has been calculated to be two-
phase;
    e. Section I.C.5.a--Post-critical heat flux correlations of heat 
transfer from the fuel cladding to the surrounding fluid; and
    f. Section I.C.7.a--Calculation of cross-flow between the hot 
and average channel regions of the core during blowdown.

VI. Issue Resolution

    A. The Commission has determined that the structures, systems, 
and components and design features of NuScale comply with the 
provisions of the Atomic Energy Act of 1954, as amended, and the 
applicable regulations identified in Section V of this appendix; and 
therefore, provide adequate protection to the health and safety of 
the public. A conclusion that a matter is resolved includes the 
finding that additional or alternative structures, systems, and 
components, design features, design criteria, testing, analyses, 
acceptance criteria, or justifications are not necessary for 
NuScale.
    B. The Commission considers the following matters resolved 
within the meaning of Sec.  52.63(a)(5) in subsequent proceedings 
for issuance of a COL, amendment of a COL, or renewal of a COL, 
proceedings held under Sec.  52.103, and enforcement proceedings 
involving plants referencing this appendix:
    1. All nuclear safety issues associated with the information in 
the final safety evaluation report, Tier 1, Tier 2, and the 
rulemaking record for certification of the NuScale design, with the 
exception of the following:
    a. generic TS and other operational requirements;
    b. the adequacy of the design of the shield wall between the 
NuScale power module and the reactor building steam gallery to limit 
potential radiological doses consistent with the radiation zones 
specified in design certification application Part 2, Tier 2, 
Chapter 12, Figure 12.3-1, ``Reactor Building Radiation Zone Map'';
    c. the adequacy of the design of the systems used for post-
accident hydrogen and oxygen monitoring described in design 
certification application Part 2, Tier 2, Section 6.2.5 to meet the 
requirements of 10 CFR 50.34(f)(2)(vii), 10 CFR 50.34(f)(2)(xxviii), 
and 10 CFR 52.47(a)(2)(iv), with respect to radiological releases 
caused by leakage from these systems under accident conditions; and
    d. the ability of the steam generator tubes to maintain 
structural and leakage integrity during density wave oscillations in 
the secondary fluid system, including the method of analysis to 
predict the thermal-hydraulic conditions of the steam generator 
secondary fluid system and resulting loads, stresses, and 
deformations from density wave oscillations and reverse flow, 
consistent with the other design information regarding steam 
generator integrity described in DCA Part 2, Tier 2, Sections 3.9.1, 
3.9.2, 5.4.1, and 15.6.3, and in accordance with 10 CFR part 50, GDC 
4 and 31;
    2. All nuclear safety and safeguards issues associated with the 
referenced information in the non-public documents in Tables 1.6-1 
and 1.6-2 of Tier 2 of the DCD, which contain sensitive unclassified 
non-safeguards information (including proprietary information and 
security-related information) and safeguards information and which, 
in context, are intended as requirements in the generic DCD for the 
NuScale design;
    3. All generic changes to the DCD under and in compliance with 
the change processes in paragraphs VIII.A.1 and VIII.B.1 of this 
appendix;
    4. All exemptions from the DCD under and in compliance with the 
change processes in paragraphs VIII.A.4 and VIII.B.4 of this 
appendix, but only for that plant;
    5. All departures from the DCD that are approved by license 
amendment, but only for that plant;
    6. Except as provided in paragraph VIII.B.5.g of this appendix, 
all departures from Tier 2 under and in compliance with the change 
processes in paragraph VIII.B.5 of this appendix that do not require 
prior NRC approval, but only for that plant; and
    7. All environmental issues concerning severe accident 
mitigation design alternatives associated with the information in 
the NRC's environmental assessment for NuScale (ADAMS Accession No. 
ML22004A006) and DCD Part 3, ``Applicant's Environmental Report--
Standard Design Certification,'' Revision 5, dated July 2020 (ADAMS 
Accession No. ML20224A512), for plants referencing this appendix 
whose site characteristics fall within the site parameters of the 
representative site specified in the NuScale environmental report.
    C. The Commission does not consider operational requirements for 
an applicant or

[[Page 3310]]

licensee who references this appendix to be matters resolved within 
the meaning of Sec.  52.63(a)(5). The Commission reserves the right 
to require operational requirements for an applicant or licensee who 
references this appendix by rule, regulation, order, or license 
condition.
    D. Except under the change processes in Section VIII of this 
appendix, the Commission may not require an applicant or licensee 
who references this appendix to:
    1. Modify structures, systems, and components or design features 
as described in the generic DCD;
    2. Provide additional or alternative structures, systems, and 
components or design features not discussed in the generic DCD; or
    3. Provide additional or alternative design criteria, testing, 
analyses, acceptance criteria, or justification for structures, 
systems, and components or design features discussed in the generic 
DCD.
    E. The NRC will specify, at an appropriate time, the procedures 
to be used by an interested person who wishes to review portions of 
the design certification or references containing safeguards 
information or sensitive unclassified non-safeguards information 
(including proprietary information, such as trade secrets and 
commercial or financial information obtained from a person that are 
privileged or confidential (10 CFR 2.390 and 10 CFR part 9), and 
security-related information), for the purpose of participating in 
the hearing required by Sec.  52.85, the hearing provided under 
Sec.  52.103, or in any other proceeding relating to this appendix, 
in which interested persons have a right to request an adjudicatory 
hearing.

VII. Duration of This Appendix

    This appendix may be referenced for a period of 15 years from 
February 21, 2023, except as provided for in Sec. Sec.  52.55(b) and 
52.57(b). This appendix remains valid for an applicant or licensee 
who references this appendix until the application is withdrawn or 
the license expires, including any period of extended operation 
under a renewed license.

VIII. Processes for Changes and Departures

A. Tier 1 Information

    1. Generic changes to Tier 1 information are governed by the 
requirements in Sec.  52.63(a)(1).
    2. Generic changes to Tier 1 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs A.3 or A.4 of this section.
    3. Departures from Tier 1 information that are required by the 
Commission through plant-specific orders are governed by the 
requirements in Sec.  52.63(a)(4).
    4. Exemptions from Tier 1 information are governed by the 
requirements in Sec. Sec.  52.63(b)(1) and 52.98(f). The Commission 
will deny a request for an exemption from Tier 1, if it finds that 
the design change will result in a significant decrease in the level 
of safety otherwise provided by the design.

B. Tier 2 Information

    1. Generic changes to Tier 2 information are governed by the 
requirements in Sec.  52.63(a)(1).
    2. Generic changes to Tier 2 information are applicable to all 
applicants or licensees who reference this appendix, except those 
for which the change has been rendered technically irrelevant by 
action taken under paragraphs B.3, B.4, or B.5, of this section.
    3. The Commission may not require new requirements on Tier 2 
information by plant-specific order, while this appendix is in 
effect under Sec.  52.55 or Sec.  52.61, unless:
    a. A modification is necessary to secure compliance with the 
Commission's regulations applicable and in effect at the time this 
appendix was approved, as set forth in Section V of this appendix, 
or to ensure adequate protection of the public health and safety or 
the common defense and security; and
    b. Special circumstances as defined in 10 CFR 50.12(a) are 
present.
    4. An applicant or licensee who references this appendix may 
request an exemption from Tier 2 information. The Commission may 
grant such a request only if it determines that the exemption will 
comply with the requirements of 10 CFR 50.12(a). The Commission will 
deny a request for an exemption from Tier 2, if it finds that the 
design change will result in a significant decrease in the level of 
safety otherwise provided by the design. The granting of an 
exemption to an applicant must be subject to litigation in the same 
manner as other issues material to the license hearing. The granting 
of an exemption to a licensee must be subject to an opportunity for 
a hearing in the same manner as license amendments.
    5.a. An applicant or licensee who references this appendix may 
depart from Tier 2 information, without prior NRC approval, unless 
the proposed departure involves a change to or departure from Tier 1 
information, or the TS, or requires a license amendment under 
paragraph B.5.b or B.5.c of this section. When evaluating the 
proposed departure, an applicant or licensee shall consider all 
matters described in the plant-specific DCD.
    b. A proposed departure from Tier 2, other than one affecting 
resolution of a severe accident issue identified in the plant-
specific DCD or one affecting information required by Sec.  
52.47(a)(28) to address aircraft impacts, requires a license 
amendment if it would:
    (1) Result in more than a minimal increase in the frequency of 
occurrence of an accident previously evaluated in the plant-specific 
DCD;
    (2) Result in more than a minimal increase in the likelihood of 
occurrence of a malfunction of a structure, system, or component 
important to safety and previously evaluated in the plant-specific 
DCD;
    (3) Result in more than a minimal increase in the consequences 
of an accident previously evaluated in the plant-specific DCD;
    (4) Result in more than a minimal increase in the consequences 
of a malfunction of a structure, system, or component important to 
safety previously evaluated in the plant-specific DCD;
    (5) Create a possibility for an accident of a different type 
than any evaluated previously in the plant-specific DCD;
    (6) Create a possibility for a malfunction of a structure, 
system, or component important to safety with a different result 
than any evaluated previously in the plant-specific DCD;
    (7) Result in a design-basis limit for a fission product barrier 
as described in the plant-specific DCD being exceeded or altered; or
    (8) Result in a departure from a method of evaluation described 
in the plant-specific DCD used in establishing the design bases or 
in the safety analyses.
    c. A proposed departure from Tier 2, affecting resolution of an 
ex-vessel severe accident design feature identified in the plant-
specific DCD, requires a license amendment if:
    (1) There is a substantial increase in the probability of an ex-
vessel severe accident such that a particular ex-vessel severe 
accident previously reviewed and determined to be not credible could 
become credible; or
    (2) There is a substantial increase in the consequences to the 
public of a particular ex-vessel severe accident previously 
reviewed.
    d. A proposed departure from Tier 2 information required by 
Sec.  52.47(a)(28) to address aircraft impacts shall consider the 
effect of the changed design feature or functional capability on the 
original aircraft impact assessment required by 10 CFR 50.150(a). 
The applicant or licensee shall describe, in the plant-specific DCD, 
how the modified design features and functional capabilities 
continue to meet the aircraft impact assessment requirements in 10 
CFR 50.150(a)(1).
    e. If a departure requires a license amendment under paragraph 
B.5.b or B.5.c of this section, it is governed by 10 CFR 50.90.
    f. A departure from Tier 2 information that is made under 
paragraph B.5 of this section does not require an exemption from 
this appendix.
    g. A party to an adjudicatory proceeding for either the 
issuance, amendment, or renewal of a license or for operation under 
Sec.  52.103(a), who believes that an applicant or licensee who 
references this appendix has not complied with paragraph VIII.B.5 of 
this appendix when departing from Tier 2 information, may petition 
to admit into the proceeding such a contention. In addition to 
complying with the general requirements of 10 CFR 2.309, the 
petition must demonstrate that the departure does not comply with 
paragraph VIII.B.5 of this appendix. Further, the petition must 
demonstrate that the change bears on an asserted noncompliance with 
an ITAAC acceptance criterion in the case of a Sec.  52.103 
preoperational hearing, or that the departure bears directly on the 
amendment request in the case of a hearing on a license amendment. 
Any other party may file a response. If, on the basis of the 
petition and any response, the presiding officer determines that a 
sufficient showing has been made, the presiding officer shall 
certify the matter directly to the Commission for determination of 
the admissibility of the contention. The Commission may admit such

[[Page 3311]]

a contention if it determines the petition raises a genuine issue of 
material fact regarding compliance with paragraph VIII.B.5 of this 
appendix.

C. Operational Requirements

    1. Changes to NuScale design certification generic TS and other 
operational requirements that were completely reviewed and approved 
in the design certification rule and do not require a change to a 
design feature in the generic DCD are governed by the requirements 
in 10 CFR 50.109. Changes that require a change to a design feature 
in the generic DCD are governed by the requirements in paragraphs A 
or B of this section.
    2. Changes to NuScale design certification generic TS and other 
operational requirements are applicable to all applicants who 
reference this appendix, except those for which the change has been 
rendered technically irrelevant by action taken under paragraphs C.3 
or C.4 of this section.
    3. The Commission may require plant-specific departures on 
generic TS and other operational requirements that were completely 
reviewed and approved, provided a change to a design feature in the 
generic DCD is not required and special circumstances, as defined in 
10 CFR 2.335 are present. The Commission may modify or supplement 
generic TS and other operational requirements that were not 
completely reviewed and approved or require additional TS and other 
operational requirements on a plant-specific basis, provided a 
change to a design feature in the generic DCD is not required.
    4. An applicant who references this appendix may request an 
exemption from the generic TS or other operational requirements. The 
Commission may grant such a request only if it determines that the 
exemption will comply with the requirements of Sec.  52.7. The 
granting of an exemption must be subject to litigation in the same 
manner as other issues material to the license hearing.
    5. A party to an adjudicatory proceeding for the issuance, 
amendment, or renewal of a license, or for operation under Sec.  
52.103(a), who believes that an operational requirement approved in 
the DCD or a TS derived from the generic TS must be changed, may 
petition to admit such a contention into the proceeding. The 
petition must comply with the general requirements of Sec.  2.309 of 
this chapter and must either demonstrate why special circumstances 
as defined in Sec.  2.335 of this chapter are present or demonstrate 
that the proposed change is necessary for compliance with the 
Commission's regulations in effect at the time this appendix was 
approved, as set forth in Section V of this appendix. Any other 
party may file a response to the petition. If, on the basis of the 
petition and any response, the presiding officer determines that a 
sufficient showing has been made, the presiding officer shall 
certify the matter directly to the Commission for determination of 
the admissibility of the contention. All other issues with respect 
to the plant-specific TS or other operational requirements are 
subject to a hearing as part of the licensing proceeding.
    6. After issuance of a license, the generic TS have no further 
effect on the plant-specific TS. Changes to the plant-specific TS 
will be treated as license amendments under 10 CFR 50.90.

IX. [Reserved]

X. Records and Reporting

A. Records

    1. The applicant for this appendix shall maintain a copy of the 
generic DCD that includes all generic changes that are made to Tier 
1 and Tier 2, and the generic TS and other operational requirements. 
The applicant shall maintain the sensitive unclassified non-
safeguards information (including proprietary information and 
security-related information) and safeguards information referenced 
in the generic DCD for the period that this appendix may be 
referenced, as specified in Section VII of this appendix.
    2. An applicant or licensee who references this appendix shall 
maintain the plant-specific DCD to accurately reflect both generic 
changes to the generic DCD and plant-specific departures made under 
Section VIII of this appendix throughout the period of application 
and for the term of the license (including any periods of renewal).
    3. An applicant or licensee who references this appendix shall 
prepare and maintain written evaluations that provide the bases for 
the determinations required by Section VIII of this appendix. These 
evaluations must be retained throughout the period of application 
and for the term of the license (including any periods of renewal).
    4.a. The applicant for NuScale shall maintain a copy of the 
aircraft impact assessment performed to comply with the requirements 
of 10 CFR 50.150(a) for the term of the certification (including any 
period of renewal).
    b. An applicant or licensee who references this appendix shall 
maintain a copy of the aircraft impact assessment performed to 
comply with the requirements of 10 CFR 50.150(a) throughout the 
pendency of the application and for the term of the license 
(including any periods of renewal).

B. Reporting

    1. An applicant or licensee who references this appendix shall 
submit a report to the NRC containing a brief description of any 
plant-specific departures from the DCD, including a summary of the 
evaluation of each departure. This report must be filed in 
accordance with the filing requirements applicable to reports in 
Sec.  52.3.
    2. An applicant or licensee who references this appendix shall 
submit updates to its plant-specific DCD, which reflect the generic 
changes to and plant-specific departures from the generic DCD made 
under Section VIII of this appendix. These updates shall be filed 
under the filing requirements applicable to final safety analysis 
report updates in 10 CFR 50.71(e) and 52.3.
    3. The reports and updates required by paragraphs X.B.1 and 
X.B.2 of this appendix must be submitted as follows:
    a. On the date that an application for a license referencing 
this appendix is submitted, the application must include the report 
and any updates to the generic DCD.
    b. During the interval from the date of application for a 
license to the date the Commission makes its finding required by 
Sec.  52.103(g), the report must be submitted semiannually. Updates 
to the plant-specific DCD must be submitted annually and may be 
submitted along with amendments to the application.
    c. After the Commission makes the finding required by Sec.  
52.103(g), the reports and updates to the plant-specific DCD must be 
submitted, along with updates to the site-specific portion of the 
final safety analysis report for the facility, at the intervals 
required by 10 CFR 50.59(d)(2) and 50.71(e)(4), respectively, or at 
shorter intervals as specified in the license.

    Dated: January 11, 2023.

    For the Nuclear Regulatory Commission.
Brooke P. Clark,
Secretary of the Commission.
[FR Doc. 2023-00729 Filed 1-18-23; 8:45 am]
BILLING CODE 7590-01-P