This section of the FEDERAL REGISTER

contains notices to the public of the proposed

issuance of rules and regulations. The

purpose of these notices is to give interested

persons an opportunity to participate in the

rule making prior to the adoption of the final

rules.

Proposed Rules Federal Register

34999 

Vol. 86, No. 124 

Thursday, July 1, 2021 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Part 52 

[NRC–2017–0029] 

RIN 3150–AJ98 

NuScale Small Modular Reactor Design 

Certification 

AGENCY: U.S. Nuclear Regulatory 

Commission. 

ACTION: Proposed rule. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC) is proposing to 

amend its regulations to certify the 

NuScale standard design for a small 

modular reactor. Applicants or licensees 

intending to construct and operate a 

NuScale standard design may do so by 

referencing this design certification rule. 

The applicant for certification of the 

NuScale standard design is NuScale 

Power, LLC. The public is invited to 

submit comments on this proposed rule. 

DATES: Submit comments by August 30, 

2021. Comments received after this date 

will be considered if it is practical to do 

so, but the NRC is able to ensure 

consideration only for comments 

received before this date. 

ADDRESSES: You may submit comments 

by any of the following methods (unless 

this document describes a different 

method for submitting comments on a 

specific subject); however, the NRC 

encourages electronic comment 

submission through the Federal 

Rulemaking website: 

• Federal Rulemaking Website: Go to 

https://www.regulations.gov and search 

for Docket ID NRC–2017–0029. Address 

questions about NRC dockets to Dawn 

Forder; telephone: 301–415–3407; 

email: Dawn.Forder@nrc.gov. For 

technical questions, contact the 

individual(s) listed in the FOR FURTHER

INFORMATION CONTACT section of this 

document. 

• Email comments to: 

Rulemaking.Comments@nrc.gov. If you 

do not receive an automatic email reply 

confirming receipt, then contact us at 

301–415–1677. 

For additional direction on obtaining 

information and submitting comments, 

see ‘‘Obtaining Information and 

Submitting Comments’’ in the 

SUPPLEMENTARY INFORMATION section of 

this document. 

FOR FURTHER INFORMATION CONTACT: 

Yanely Malave, Office of Nuclear 

Material Safety and Safeguards, 

telephone: 301–415–1519, email: 

Yanely.Malave@nrc.gov, and Prosanta 

Chowdhury, Office of Nuclear Reactor 

Regulation, telephone: 301–415–1647, 

email: Prosanta.Chowdhury@nrc.gov. 

Both are staff of the U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001. 

SUPPLEMENTARY INFORMATION: 

Table of Contents 

I. Obtaining Information and Submitting 

Comments 

II. Background 

III. Regulatory and Policy Issues 

IV. Technical Issues Associated With the 

NuScale Design 

V. Discussion 

A. Introduction (Section I) 

B. Definitions (Section II) 

C. Scope and Contents (Section III) 

D. Additional Requirements and 

Restrictions (Section IV) 

E. Applicable Regulations (Section V) 

F. Issue Resolution (Section VI) 

G. Duration of This Appendix (Section VII) 

H. Processes for Changes and Departures 

(Section VIII) 

I. [Reserved] (Section IX) 

J. Records and Reporting (Section X) 

VI. Section-by-Section Analysis 

VII. Regulatory Flexibility Certification 

VIII. Regulatory Analysis 

IX. Backfitting and Issue Finality 

X. Plain Writing 

XI. Environmental Assessment and Finding 

of No Significant Impact 

XII. Paperwork Reduction Act 

XIII. Agreement State Compatibility 

XIV. Voluntary Consensus Standards 

XV. Availability of Documents 

XVI. Procedures for Access to Proprietary 

and Safeguards Information for 

Preparation of Comments on the NuScale 

Design Certification Proposed Rule 

XVII. Incorporation by Reference— 

Reasonable Availability to Interested 

Parties 

I. Obtaining Information and 

Submitting Comments 

A. Obtaining Information 

Please refer to Docket ID NRC–2017– 

0029 when contacting the NRC about 

the availability of information for this 

proposed rule. You may obtain publicly 

available information related to this 

proposed rule by any of the following 

methods: 

• Federal Rulemaking Website: Go to 

https://www.regulations.gov and search 

for Docket ID NRC–2017–0029. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may obtain publicly 

available documents online in the 

ADAMS Public Documents collection at 

https://www.nrc.gov/reading-rm/ 

adams.html. To begin the search, select 

‘‘Begin Web-based ADAMS Search.’’ For 

problems with ADAMS, please contact 

the NRC’s Public Document Room (PDR) 

reference staff at 1–800–397–4209, at 

301–415–4737, or by email to 

PDR.Resource@nrc.gov. The ADAMS 

accession number for each document 

referenced in this proposed rule (if that 

document is available in ADAMS) is 

provided the first time that it is 

mentioned in this document. In 

addition, for the convenience of the 

reader, instructions about obtaining 

materials referenced in this document 

are provided in Section XV, 

‘‘Availability of Documents,’’ of this 

document. 

• Attention: The Public Document 

Room (PDR), where you may examine 

and order copies of public documents, 

is currently closed. You may submit 

your request to the PDR via email at 

PDR.Resource@nrc.gov or by calling 

1–800–397–4209 between 8:00 a.m. and 

4:00 p.m. (ET), Monday through Friday, 

except Federal holidays. 

• Attention: The Technical Library, 

which is located at Two White Flint 

North, 11545 Rockville Pike, Rockville, 

Maryland 20852, is open by 

appointment only. Interested parties 

may make appointments to examine 

documents by contacting the NRC 

Technical Library by email at 

Library.Resource@nrc.gov between 8:00 

a.m. and 4:00 p.m. (ET), Monday 

through Friday, except Federal holidays. 

B. Submitting Comments 

The NRC encourages electronic 

comment submission through the 

Federal Rulemaking website (https://

www.regulations.gov). Please include 

Docket ID NRC–2017–0029 in your 

comment submission. 

The NRC cautions you not to include 

identifying or contact information that 

you do not want to be publicly 


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disclosed in your comment submission. 

The NRC will post all comment 

submissions at https://

www.regulations.gov as well as enter the 

comment submissions into ADAMS. 

The NRC does not routinely edit 

comment submissions to remove 

identifying or contact information. 

If you are requesting or aggregating 

comments from other persons for 

submission to the NRC, then you should 

inform those persons not to include 

identifying or contact information that 

they do not want to be publicly 

disclosed in their comment submission. 

Your request should state that the NRC 

does not routinely edit comment 

submissions to remove such information 

before making the comment 

submissions available to the public or 

entering the comment submissions into 

ADAMS. 

II. Background 

Part 52 of title 10 of the Code of 

Federal Regulations (10 CFR), 

‘‘Licenses, Certifications, and Approvals 

for Nuclear Power Plants,’’ subpart B, 

‘‘Standard Design Certifications,’’ 

presents the process for obtaining 

standard design certifications. By letter 

dated December 31, 2016, NuScale 

Power, LLC, (NuScale Power) filed its 

application for certification of the 

NuScale standard design (hereafter 

referred to as NuScale) (ADAMS 

Accession No. ML17013A229). The NRC 

published a notification of receipt of the 

design certification application (DCA) in 

the Federal Register on February 22, 

2017 (82 FR 11372). On March 30, 2017, 

the NRC published a notification of 

acceptance for docketing of the 

application in the Federal Register (82 

FR 15717) and assigned docket number 

52–048. The preapplication information 

submitted before the NRC formally 

accepted the application can be found 

in ADAMS under Docket No. PROJ0769. 

NuScale is the first small modular 

reactor design reviewed by the NRC. 

NuScale is based on a small light water 

reactor developed at Oregon State 

University in the early 2000s. It consists 

of one or more NuScale power modules 

(hereafter referred to as power 

module(s)). A power module is a natural 

circulation light water reactor composed 

of a reactor core, a pressurizer, and two 

helical coil steam generators located in 

a common reactor pressure vessel that is 

housed in a compact cylindrical steel 

containment. The NuScale reactor 

building is designed to hold up to 12 

power modules. Each power module has 

a rated thermal output of 160 megawatt 

thermal (MWt) and electrical output of 

50 megawatt electric (MWe), yielding a 

total capacity of 600 MWe for 12 power 

modules. All NuScale power modules 

are partially submerged in one safetyrelated pool, which is also the ultimate 

heat sink for the reactor. The pool 

portion of the reactor building is located 

below grade. The design utilizes several 

first-of-a-kind approaches for 

accomplishing key safety functions, 

resulting in no need for Class 1E safetyrelated power (no emergency diesel 

generators), no need for pumps to inject 

water into the core for post-accident 

coolant injection, and reduced need for 

control room staffing while providing 

safe operation of the plant during 

normal and post-accident operation. 

III. Regulatory and Policy Issues 

A. Control Room Staffing Requirements 

The requirements in § 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 § 52.103(g) that 

the acceptance criteria in the COL are 

met. 

In a letter to the NRC, dated 

September 15, 2015 (ADAMS Accession 

No. ML15258A846), NuScale Power 

proposed that 6 licensed operators 

would operate up to 12 power modules 

from a single control room. The staffing 

proposal would meet the requirements 

of § 50.54(k) but would not meet the 

requirements in § 50.54(m)(2)(i) because 

the minimum requirements for the 

onsite staffing table in § 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 

§ 50.54(m)(2)(iii), which requires a 

licensed operator at the controls for 

each fueled unit (i.e., 12 licensed 

operators). Absent alternative staffing 

requirements, future applicants 

referencing the NuScale design would 

need to request an exemption. 

In the DCA Part 7, Section 6.2, 

‘‘Justification for Rulemaking,’’ NuScale 

Power provided a technical basis for 

rulemaking language that would address 

control room staffing in conjunction 

with control room configuration. 

NuScale Power’s approach is consistent 

with SECY–11–0098, ‘‘Operator Staffing 

for Small or Multi-Module Nuclear 

Power Plant Facilities,’’ dated July 22, 

2011 (ADAMS Accession No. 

ML111870574). In Chapter 18, Section 

18.5.4.2, ‘‘Evaluation of the Applicant’s 

Technical Basis,’’ of the final safety 

evaluation report (ADAMS Accession 

No. ML20023B605), the NRC found that 

NuScale Power’s proposed staffing 

level, as described in the DCA Part 7, 

Section 6, is acceptable. Because 

Section V, ‘‘Applicable Regulations,’’ of 

this proposed rule includes the 

alternative staffing requirement 

provisions, staffing table, and 

appropriate table notes, a future 

applicant or licensee that references 

proposed appendix G to 10 CFR part 52 

would not need to request an exemption 

from § 50.54(m). 

B. Incorporation by Reference 

The proposed Section III.A, 

‘‘Incorporation by reference approval,’’ 

of appendix G to 10 CFR part 52 lists 

documents that would be approved by 

the Director of the Office of the Federal 

Register for incorporation by reference 

into this appendix. Proposed Section 

III.B.2 identifies information that is not 

within the scope of the design 

certification and, therefore, is not 

incorporated by reference into this 

appendix. This information includes 

conceptual design information, as 

defined in § 52.47(a)(24), and the 

discussion of ‘‘first principles’’ 

described in the Design Control 

Document (DCD) Part 2, Tier 2, Section 

14.3.2, ‘‘Tier 1 Design Description and 

Inspections, Tests, Analyses, and 

Acceptance Criteria First Principles.’’ 

C. Issues Not Resolved by the Design 

Certification 

The NRC identified three issues as not 

resolved within the meaning of 

§ 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 thermalhydraulic conditions of the steam 

generator secondary fluid system and 

resulting loads, stresses, and 

deformations from density wave 

oscillations from reverse flow. 

1. Shielding Wall Design 

As discussed in Section 12.3.4.1.2 of 

the final safety evaluation report, the 

NRC found that there were insufficient 

design details available regarding 

shielding wall design with the presence 

of large penetrations, such as the main 


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steam lines; main feedwater lines; and 

power module bay heating, ventilation, 

and air conditioning lines in the 

radiation shield wall between the power 

module bay and the reactor building 

steam gallery area. Without this 

shielding design information, the NRC 

is unable to confirm that the 

radiological doses to workers will be 

maintained within the radiation zone 

limits specified in the application. 

This issue is narrowly focused on the 

shielding walls between the reactor 

module bays and the reactor building 

steam gallery areas. The radiation zones 

and dose calculations, including dose 

calculations for the dose to workers, 

members of the public, and 

environmental qualification, in areas 

outside of the reactor module bay are 

calculated assuming a solid wall and 

currently do not account for 

penetrations in the shield wall. A COL 

applicant would be required to 

demonstrate penetration shielding 

adequate to address the following issues 

in the NuScale DCD: The plant radiation 

zones, environmental qualification dose 

calculations, and dose estimates for 

workers and the public. A COL 

applicant can provide this information 

for the NRC to review because this issue 

involves a localized area of the plant 

without affecting other aspects of the 

NRC’s review of the NuScale design. 

Therefore, the NRC has determined that 

this information can be provided by a 

COL applicant that references this 

appendix without a demonstrable 

impact on safety or standardization. 

Appendix G to 10 CFR part 52, Section 

VI, ‘‘Issue Resolution,’’ would clarify 

that this issue is not resolved within the 

meaning of § 52.63(a)(5), and Section IV, 

‘‘Additional Requirements and 

Restrictions,’’ would state that the COL 

applicant is responsible for providing 

the design information to address this 

issue. 

2. Containment Leakage From the 

Combustible Gas Monitoring System 

As documented in Section 12.3.4.1.3 

of the final safety evaluation report, 

there was insufficient information 

available regarding NuScale 

combustible gas monitoring system and 

the potential for leakage from this 

system outside containment. Without 

additional information regarding the 

potential for leakage from this system, 

the NRC was unable to determine 

whether this leakage could impact 

analyses performed to assess main 

control room dose consequences, offsite 

dose consequences to members of the 

public, and whether this system can be 

safely re-isolated after monitoring is 

initiated due to potentially high dose 

levels at or near the isolation valve 

location. The isolation valve can only be 

operated locally, and dose levels at the 

valve location have not been 

determined. 

This issue is narrowly focused on the 

radiation dose implications as a result of 

using the post-accident combustible gas 

monitoring loop. A COL applicant 

would be required to demonstrate either 

that offsite and main control room dose 

calculations are not exceeded or that the 

system can be safely re-isolated, if 

needed. This issue does not affect 

normal plant operation or non-core 

damage accidents. The issue may be 

resolved by performing radiation dose 

calculations and demonstrating that 

doses would remain within applicable 

dose limits in 10 CFR part 20, 

‘‘Standards for Protection Against 

Radiation.’’ More information may be 

available at the COL application stage 

that would allow for more detailed 

calculations. Any design changes to 

address this issue would only affect the 

combustible gas monitoring loop to 

ensure it can be re-isolated or to ensure 

that dose limits are not exceeded. Such 

design changes would likely not have an 

impact on other systems or equipment, 

and the NRC would review such 

changes and any resulting effects on 

other structures, systems, and 

components during the COL application 

review to provide reasonable assurance 

of adequate protection. Therefore, the 

NRC has determined that this 

information can be provided by a COL 

applicant that references this appendix 

without a demonstrable impact on 

safety or standardization. Appendix G to 

10 CFR part 52, Section VI, ‘‘Issue 

Resolution,’’ would clarify that this 

issue is not resolved within the meaning 

of § 52.63(a)(5), and Section IV, 

‘‘Additional Requirements and 

Restrictions,’’ would state that the COL 

applicant is responsible for providing 

the design information to address this 

issue. 

3. Steam Generator Stability During 

Density Wave Oscillations and 

Associated Method of Analysis 

Section 5.4.1.2, ‘‘System Design,’’ in 

Revision 2 of the DCA Part 2, Tier 2, 

stated that a flow restriction device at 

the inlet to each steam generator tube 

‘‘ensures secondary-side flow stability 

and precludes density wave 

oscillations.’’ However, the applicant 

modified this section in Revision 3 of 

the DCA Part 2, Tier 2 to state that the 

steam generator inlet flow restrictors 

provide the necessary secondary-side 

pressure drop ‘‘to reduce flow 

oscillations to acceptable limits.’’ 

Revision 4.1 of the DCA (ADAMS 

Accession No. ML20205L562) revised 

Section 5.4.1.2 to state that the steam 

generator inlet flow restrictors are 

designed ‘‘to reduce the potential for 

density wave oscillations.’’ Revision 5 

of the DCA (ADAMS Accession No. 

ML20225A071) provides only editorial 

changes to Revision 4.1 and does not 

change the technical content or 

conclusions. 

Sections 3.9.2, 3.9.5, and 5.4.1 of the 

final safety evaluation report relied on 

the applicant’s statements in Revision 2 

and Revision 3 of the DCA that flow 

oscillations in the secondary fluid 

system of the steam generators would 

either be precluded or minimal. After 

issuance of the advanced safety 

evaluation report, the NRC noted 

inconsistencies and gaps in the 

information provided in Sections 3.9.1, 

3.9.2, and 5.4.1 of Revision 4.1 of the 

DCA Part 2, Tier 2 regarding the 

potential for significant density wave 

oscillations in the steam generator 

tubes, including both forward and 

reverse secondary flow. The testing 

performed by the applicant on various 

conceptual designs of the steam 

generator inlet flow restrictors only 

involved flow in the forward direction 

without oscillation or reverse flow. 

As a result, NuScale Power has not 

demonstrated that the flow oscillations 

that are predicted to occur on the 

secondary-side of the steam generators 

will not cause failure of the inlet flow 

restrictors. Structural and leakage 

integrity of the inlet flow restrictors in 

the steam generators is necessary to 

avoid damage to multiple steam 

generator tubes, caused directly by 

broken parts or indirectly by 

unexpected density wave oscillation 

loads. Damage to multiple steam 

generator tubes could disrupt natural 

circulation in the reactor coolant 

pathway and interfere with the decay 

heat removal system and the emergency 

core cooling system, which is relied 

upon to cool the reactor core in a 

NuScale nuclear power module. The 

failure of multiple steam generator tubes 

resulting from failure of an inlet flow 

restrictor has not been included within 

the scope of the NuScale accident 

analyses in DCA Part 2, Tier 2, Chapter 

15. Therefore, the NRC concludes that 

NuScale Power has not demonstrated 

compliance with 10 CFR part 20 and 10 

CFR part 50, appendix A, General 

Design Criterion (GDC) 4 and GDC 31, 

relative to potential impacts on steam 

generator tube integrity from inlet flow 

restrictor failure. 

As described previously, NuScale 

Power made a change to the description 

of inlet flow restrictor performance 

beginning with DCA Part 2, Tier 2, 


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Revision 3, that indicates that the design 

no longer precludes density wave 

oscillations in the secondary-side of the 

steam generators. As a result, the design 

needs a method of analysis to predict 

the thermal-hydraulic conditions of the 

steam generator secondary fluid system 

and resulting loads, stresses, and 

deformations from density wave 

oscillations including reverse flow. 

However, an appropriate method of 

analysis has not been provided to the 

NRC. 

The DCA Part 2, Tier 2, Section 

3.9.1.2, ‘‘Computer Programs Used in 

Analyses,’’ lists the computer programs 

used by NuScale Power in the dynamic 

and static analyses of mechanical loads, 

stresses, and deformations, and in the 

hydraulic transient load analyses of 

seismic Category I components and 

supports for the NuScale nuclear power 

plant. Section 3.9.1.2 states that 

NRELAP5 is NuScale’s proprietary 

system thermal-hydraulics code for use 

in safety-related design and analysis 

calculations and is pre-verified and 

configuration-managed. The advanced 

safety evaluation report, Section 

3.9.1.4.9, ‘‘Computer Programs Used in 

Analyses,’’ states that the NRELAP5 

computer program had received 

verification and validation. Following 

preparation of the advanced safety 

evaluation report, the NRC noted a 

discrepancy between two statements in 

the DCA about validation for NRELAP5: 

DCA Part 2, Tier 2, Section 5.4.1.3 in 

Revision 4 stated that NRELAP5 was 

validated for determining density wave 

oscillation thermal-hydraulic 

conditions, referring to Section 15.0.2 

for more information, but neither 

Section 15.0.2 nor TR–1016–51669 

describe validation for determining 

density wave oscillation thermalhydraulic conditions. 

On June 19, 2020, NuScale submitted 

Revision 4.1 of the DCA Part 2, Tier 2 

(ADAMS Accession No. ML20205L562; 

subsequently included in Revision 5 of 

the DCA submitted on July 29, 2020 

(ADAMS Accession No. 

ML20225A071)) to correct the 

discrepancies, and acknowledges the 

need for a COL applicant to address 

secondary-side instabilities in the steam 

generator design. Specifically, the 

update to Section 3.9.1.2 in Revision 4.1 

of DCA Part 2, Tier 2, references DCA 

Part 2, Tier 2, Section 15.0.2, ‘‘Review 

of Transient and Accident Analysis 

Methods,’’ for the discussion of the 

development, use, verification, 

validation, and code limitations of the 

NRELAP5 computer program for 

application to transient and accident 

analyses. The correction to Section 

3.9.1.2 also references technical report 

TR–1016–51669, ‘‘NuScale Power 

Module Short-Term Transient 

Analysis,’’ incorporated by reference in 

DCA Part 2, Tier 2, Table 1.6–2, for 

application of the NRELAP5 computer 

program to short-term transient dynamic 

mechanical loads, such as pipe breaks 

and valve actuations. In addition, the 

correction to Section 3.9.1.2 includes a 

new COL item specifying that a COL 

applicant that references the NuScale 

DCD would develop an evaluation 

methodology for the analysis of 

secondary-side instabilities in the steam 

generator design. The COL item states 

that this methodology would address 

the identification of potential density 

wave oscillations in the steam generator 

tubes and qualification of the applicable 

portions of the reactor coolant system 

integral reactor pressure vessel and 

steam generator given the occurrence of 

density wave oscillations, including the 

effects of reverse fluid flows within the 

tubes. These corrections to the DCA 

clarify that the evaluation methodology 

for the analysis of secondary-side 

instabilities in the steam generator 

design was not verified and validated as 

part of the NuScale DCA but would be 

accomplished by the COL applicant. 

This steam generator design issue is 

narrowly focused on the effects of 

density wave oscillations in the 

secondary fluid system on steam 

generator tubes to maintain structural 

and leakage integrity, including the 

method of analysis to predict the 

thermal-hydraulic conditions of the 

steam generator secondary fluid system 

and resulting loads, stresses, and 

deformations from density wave 

oscillations including reverse flow. No 

other reactor safety aspect of the steam 

generators is impacted by this design 

issue. As a result, the NRC finds that 

this is an isolated issue that does not 

affect other aspects of the NRC’s review 

of the design of the NuScale nuclear 

power plant. Therefore, the NRC has 

determined that this information can be 

provided by a COL applicant that 

references this appendix, consistent 

with the other design information 

regarding steam generator integrity 

described in DCA Part 2, Tier 2, 

Sections 3.9.1, 3.9.2, and 5.4.1, without 

a demonstrable impact on safety or 

standardization. Therefore, appendix G 

to 10 CFR part 52, Section VI, ‘‘Issue 

Resolution,’’ would clarify that this 

issue is not resolved within the meaning 

of § 52.63(a)(5), and Section IV, 

‘‘Additional Requirements and 

Restrictions,’’ would state that the COL 

applicant is responsible for providing 

the design information to address this 

issue. 

IV. Technical Issues Associated With 

the NuScale Design 

The NRC identified significant 

technical issues associated with the 

following design areas that were 

resolved by NuScale Power during the 

review: 

• Comprehensive vibration 

assessment program; 

• Containment safety analysis; 

• Emergency core cooling system 

inadvertent actuation block valve; 

• Conformance with GDC 27, 

‘‘Combined Reactivity Control Systems 

Capability,’’ of appendix A, ‘‘General 

Design Criteria for Nuclear Power 

Plants,’’ to 10 CFR part 50; 

• Absence of safety-related Class 1E 

alternating current (AC) or direct 

current (DC) electrical power; 

• Accident source term methodology; 

• Boron redistribution during passive 

cooling modes. 

In addition, the NRC granted 17 

exemptions from 10 CFR part 50 to 

address various aspects of NuScale’s 

design. 

A. Comprehensive Vibration 

Assessment Program 

The NuScale comprehensive vibration 

assessment program limits potentially 

adverse effects from flow, acoustic, and 

mechanically induced vibrations and 

resonances on NuScale power module 

components, including the helical coil 

steam generators. The NuScale steam 

generators are different from those of 

operating pressurized-water reactors in 

that the primary reactor coolant is on 

the outside of the steam generator tubes 

and the steam is on the inside. Because 

of this design, there is the possibility of 

density wave oscillation instabilities in 

the secondary coolant which could 

challenge the integrity of the tubes. The 

NRC’s review and findings, including 

independent analyses and observation 

of vibration testing, are documented in 

detail in Chapter 3, ‘‘Design of 

Structures, Components, Equipment, 

and Systems,’’ Section 3.9.2, ‘‘Dynamic 

Testing and Analysis of Systems, 

Structures, and Components,’’ of the 

final safety evaluation report. The 

review focused on assuring that the 

design of the helical coil steam 

generator tubes would not result in 

issues with flow-induced vibration. 

As part of the comprehensive 

vibration assessment, the NRC also 

reviewed and found acceptable the 

steam generator tube margin against 

fluid-elastic instability, steam generator 

tube margin against vortex shedding, 

control rod drive shaft margin against 

vortex shedding, in-core instrument 

guide tube against vortex shedding, 


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decay heat removal system piping 

against acoustic resonance, and control 

rod assembly guide tube against 

turbulence buffeting. The steam 

generator tube margins against fluidelastic instability and vortex shedding 

will be validated in the TF–3 testing 

facility as described in DCA Part 2, Tier 

1, Section 2.1.1, ‘‘Design Description.’’ 

In addition, the initial startup testing 

will confirm that flow-induced vibration 

will not cause adverse effects on the 

plant system components including the 

steam generator tubes. With the 

exception of the steam generator tube 

and inlet flow restrictor issue discussed 

previously, the NRC found the 

comprehensive vibration assessment 

program adequate to ensure the 

structural integrity of the NuScale 

power module components. 

B. Containment Safety Analysis 

NuScale incorporates novel and 

unique features which result in 

transient thermal-hydraulic responses 

that are different from those of currently 

licensed reactors. 

There are several peak containment 

pressure analysis technical issues 

unique to NuScale, including the 

associated thermal-hydraulic analyses. 

In support of containment safety 

analysis, NuScale Power submitted 

technical report TR–0516–49084–P, 

Revision 3, ‘‘Containment Response 

Analysis Methodology,’’ May 2020 

(ADAMS Accession No. ML20141L808) 

that describes the conservative 

containment pressure and temperature 

safety analyses for several design-basis 

events related to the containment design 

margins. NuScale also submitted topical 

report TR–0516–49422, ‘‘Loss-ofCoolant Accident Evaluation Model,’’ 

Revision 1, dated November 2019 

(ADAMS Accession No. ML19331B585). 

This topical report describes the 

evaluation model used to analyze the 

power module response during a 

design-basis loss-of-coolant accident. 

The NRC reviewed this topical report as 

part of the containment safety analysis. 

The NRC also observed thermalhydraulic performance testing at 

NuScale Power’s integrated system test 

facility, which validates the analytical 

model. Based on initial testing results 

and thermal-hydraulic analyses, 

NuScale Power made design changes to 

increase the initial reactor building pool 

level and the in-containment vessel 

design pressure to account for some 

uncertainties. 

The NRC reviewed the details of the 

computer thermal-hydraulic evaluation 

model described in the DCA Part 2, Tier 

2, Section 6.2.1.1 to determine whether 

any uncertainties were properly 

accounted for and found the 

containment design margins to be 

acceptable. The associated safety 

evaluation report approving topical 

report TR–0516–49422 was issued on 

February 18, 2020 (ADAMS Accession 

No. ML20044E199). The NRC’s review 

and specific findings, including 

independent analyses and observation 

of NuScale testing, are documented in 

Chapter 6, ‘‘Engineered Safety 

Features,’’ Section 6.2.1.1, 

‘‘Containment Structure,’’ of the safety 

evaluation report. 

C. Emergency Core Cooling System 

Inadvertent Actuation Block Valve 

The NuScale emergency core cooling 

system relies on natural circulation 

cooling of the reactor core by releasing 

the heated reactor coolant steam from 

the top of the reactor pressure vessel 

through three reactor vent valves into 

the containment vessel and returning 

the cooled condensed reactor coolant 

water to the reactor pressure vessel 

through two reactor recirculation valves. 

Each reactor vent valve and reactor 

recirculation valve consists of a first-ofa-kind arrangement of a main valve, an 

inadvertent actuation block (IAB) valve, 

a solenoid trip valve, and a solenoid 

reset valve. The IAB valve for each 

reactor vent valve and reactor 

recirculation valve is designed to close 

rapidly to prevent its corresponding 

emergency core cooling system main 

valve from opening when the reactor 

coolant system is at high pressure 

conditions. Premature opening of the 

emergency core cooling system main 

valves could result in fuel damage. The 

IAB valve then opens at reduced reactor 

coolant system pressure to allow the 

main valve to open and permit natural 

circulation cooling of the reactor core in 

response to a plant event. Although the 

valve assemblies are considered an 

active component, NuScale does not 

apply the single failure criterion to the 

IAB valve, including to the IAB valve’s 

function to close. Consistent with 

Commission safety goals and the 

practice of risk-informed 

decisionmaking, the NRC evaluated the 

NuScale emergency core cooling system 

valve system without assuming a single 

active failure of the IAB valve to close. 

During design demonstration tests of 

the first-of-a-kind emergency core 

cooling system valve system performed 

under § 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 

§ 50.55a require that a NuScale nuclear 

power plant satisfy American Society of 

Mechanical Engineers Operation and 

Maintenance of Nuclear Power Plants, 

Division 1, OM Code: Section IST (OM 

Code) as incorporated by reference in 

§ 50.55a for inservice testing of the 

emergency core cooling system valves, 

unless relief is granted or an alternative 

is authorized by the NRC. The NRC’s 

review and findings related to the IAB 

valve are documented in safety 

evaluation report Chapter 3, ‘‘Design of 

Structures, Components, Equipment, 

and Systems,’’ Section 3.9.6, 

‘‘Functional Design, Qualification, and 

Inservice Testing Programs for Pumps, 

Valves, and Dynamic Restraints.’’ These 

findings show that the NRC regulatory 

requirements and DCD Part 2, Tier 2 

provisions provide reasonable assurance 

that the emergency core system valve 

system will be capable of performing its 

design-basis functions in light of the 

safety significance of the required 

opening and closing pressures for the 

individual IAB valves. 

Further, Chapter 15, ‘‘Transient and 

Accident Analyses,’’ Section 15.0.0.5, 

‘‘Limiting Single Failures,’’ of the safety 

evaluation report states that the IAB 

valve is a first-of-a-kind, safetysignificant, active component integral to 

the NuScale emergency core cooling 

system. NuScale does not apply the 

single failure criterion to the IAB valve, 

and the Commission directed the staff in 

SRM–SECY–19–0036, ‘‘Staff 

Requirements—SECY–19–0036— 

Application of the Single Failure 

Criterion to NuScale Power LLC’s 

Inadvertent Actuation Block Valves,’’ 

(ADAMS Accession No. ML19183A408) 

to ‘‘review Chapter 15 of the NuScale 

Design Certification Application 

without assuming a single active failure 

of the inadvertent actuation block valve 

to close.’’ The Commission further 

stated that ‘‘[t]his approach is consistent 

with the Commission’s safety goal 

policy and associated core damage and 

large release frequency goals and 

existing Commission direction on the 

use of risk-informed decision-making, as 

articulated in the 1995 Policy Statement 


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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 RiskInformed and Performance-Based 

Regulation,’’ and Yellow 

Announcement 99–019).’’ 

Based on the NRC’s historic 

application of the single failure criterion 

and Commission direction on the 

subject, as described in SECY–77–439, 

‘‘Single Failure Criterion’’ (ADAMS 

Accession No. ML060260236), SRM– 

SECY–94–084, ‘‘Policy and Technical 

Issues associated with the Regulatory 

Treatment of Non-Safety Systems and 

Implementation of Design Certification 

and Light-Water Reactor Design Issues’’ 

(ADAMS Accession No. ML003708098), 

and SRM–SECY–19–0036, the NRC has 

retained discretion, in fact- or 

application-specific circumstances, to 

decide when to apply the single failure 

criterion. The Commission’s decision in 

SRM–SECY–19–0036 provides direction 

regarding the appropriate application 

and interpretation of the regulatory 

requirements in 10 CFR part 50 to the 

NuScale IAB valve’s function to close. 

This decision is similar to those in 

previous Commission documents that 

addressed the use of the single failure 

criterion and provided clarification on 

when to apply the single failure 

criterion in other specific instances. 

D. Exemption to General Design 

Criterion 27, ‘‘Combined Reactivity 

Control Systems Capability’’ 

NuScale Power determined that, 

under certain end-of-cycle scenarios 

with one control rod stuck out, the 

NuScale reactivity control systems 

could not prevent re-criticality and 

return to power. This result does not 

meet GDC 27 of appendix A to 10 CFR 

part 50, which covers reactivity control 

systems to reliably control reactivity 

changes under postulated accident 

conditions with margin for stuck control 

rods. Therefore, NuScale Power 

submitted an exemption request for 

GDC 27 (refer to Section 15, ‘‘10 CFR 50, 

Appendix A, Criterion 27, Combined 

Reactivity Control Systems Capability,’’ 

of DCA Part 7, ‘‘Exemptions’’). 

NuScale Power analyses determined 

that the specified acceptable fuel design 

limits would not be exceeded and that 

core cooling would be maintained 

during a return to power under these 

scenarios. The global core power level 

would be less than 10 percent and 

within capacity of the safety-related, 

passive decay heat removal system. The 

NRC independently verified NuScale 

Power’s results and found that NuScale 

achieves the fundamental safety 

functions for nuclear reactor safety, 

which are to control heat generation, 

remove heat, and limit the release of 

radioactive materials. Chapter 15, 

Section 15.0.6.4.1, of the safety 

evaluation report contains details of the 

evaluation of this exemption request. 

Additional information is provided in 

SECY–18–0099, ‘‘NuScale Power 

Exemption Request from 10 CFR part 

50, Appendix A, General Design 

Criterion 27, ‘Combined Reactivity 

Control Systems Capability’’’ (ADAMS 

Accession No. ML18065A431), dated 

October 9, 2018. The NRC granted the 

exemption request. 

E. Safety-Related Class 1E AC or DC 

Electrical Power 

NuScale does not contain safetyrelated 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 safetyrelated electrical power or external 

sources of coolant inventory makeup. 

NuScale Power provided a methodology 

to substantiate its assertion that the 

safety-related systems do not rely on 

Class 1E electrical power in topical 

report TR–0815–16497, ‘‘Safety 

Classification of Passive Nuclear Power 

Plant Electrical Systems,’’ dated 

February 23, 2018 (ADAMS Accession 

No. ML18054B607). The NRC reviewed 

topical report TR–0815–16497 and 

concluded that NuScale Power 

demonstrated that the safety-related 

systems do not rely on Class 1E 

electrical power. The NRC’s review and 

conclusions are documented in a safety 

evaluation report approving topical 

report TR–0815–16497 (ADAMS 

Accession No. ML17048A459) issued 

December 13, 2017, as described in the 

final safety evaluation report for Chapter 

1, ‘‘Introduction and General 

Discussion,’’ (ADAMS Accession No. 

ML20204A986). 

Because no safety-related functions of 

NuScale rely on electrical power, 

NuScale does not need any safetyrelated electrical power systems. 

Therefore, NuScale Power requested an 

exemption from GDC 17, which requires 

the provision of onsite and offsite power 

to provide sufficient capacity and 

capability to assure that (1) specified 

acceptable fuel design limits and design 

conditions of the reactor coolant 

pressure boundary are not exceeded as 

a result of anticipated operational 

occurrences and (2) the core is cooled 

and containment integrity and other 

vital functions are maintained in the 

event of postulated accidents. The NRC 

determined that, subject to limitations 

and conditions stipulated in its safety 

evaluation report for TR–0815–16497, 

the underlying purpose of GDC 17 (to 

ensure sufficient electric power is 

available to accomplish the safety 

functions of the respective systems), is 

met without reliance on Class 1E 

electric power. In other words, the 

onsite and offsite electric power systems 

are classified as non-Class 1E systems 

and electric power is not needed (1) to 

achieve or maintain safe shutdown, (2) 

to assure specified acceptable fuel 

design limits and design conditions of 

the reactor coolant pressure boundary 

are not exceeded as a result of 

anticipated operational occurrences, or 

(3) to maintain core cooling, 

containment integrity, and other vital 

functions during postulated accidents. 

Further, the onsite and offsite power 

systems are not needed to permit 

functioning of structures, systems, and 

components important to safety. 

Therefore, NuScale Power was granted 

an exemption from GDC 17. The NRC’s 

evaluation of NuScale Power’s 

exemption request from the 

requirements of GDC 17 is documented 

in Section 8.1.5, ‘‘Technical Evaluation 

for Exemptions,’’ of the final safety 

evaluation report for Chapter 8, 

‘‘Electric Power’’ (ADAMS Accession 

No. ML20023B614). 

F. Accident Source Term Methodology 

The NRC reviewed NuScale Power’s 

methods for developing accident source 

terms and performing accident 

radiological consequence analyses. As 

defined in § 50.2, ‘‘Definitions,’’ a 

source term ‘‘refers to the magnitude 

and mix of the radionuclides released 

from the fuel, expressed as fractions of 

the fission product inventory in the fuel, 

as well as their physical and chemical 

form, and the timing of their release.’’ 

NuScale Power developed source terms 

for deterministic accidents for NuScale 

that are similar to those which have 

been used in safety and siting 

assessments for large light water 

reactors. The design-basis accidents for 


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NuScale are the main steam line break 

outside containment, rod ejection 

accident, fuel handling accident, steam 

generator tube failure, and the failure of 

small lines carrying primary coolant 

outside containment. 

To address the source term regulatory 

requirements, NuScale Power submitted 

topical report TR–0915–17565, Revision 

3, ‘‘Accident Source Term 

Methodology,’’ dated April 2019 

(ADAMS Accession No. ML19112A172). 

The topical report proposes a 

methodology to develop a source term 

based on several severe accident 

scenarios that result in core damage, 

taken from the design probabilistic risk 

assessment. This source term is the 

surrogate radiological source term for a 

core damage event. 

The topical report also provides 

methods for determining radiation 

sources not developed from core 

damage scenarios for use in the 

evaluation of environmental 

qualification of equipment under 

§ 50.49, ‘‘Environmental qualification of 

electric equipment important to safety 

for nuclear power plants.’’ Specifically, 

the report describes an iodine spike 

source term not involving core damage, 

which is a surrogate accident that 

bounds potential accidents with release 

of the reactor coolant into the 

containment vessel. 

The staff submitted a related 

information paper to the Commission, 

SECY–19–0079, ‘‘Staff Approach to 

Evaluate Accident Source Terms for the 

NuScale Power Design Certification 

Application,’’ dated August 16, 2019 

(ADAMS Accession No. ML19107A455), 

describing the regulatory and technical 

issues raised by unique aspects of 

NuScale Power’s proposed methodology 

and the staff’s approach to reviewing 

topical report TR–0915–17565. 

The NRC’s review and findings of 

topical report TR–0915–17565, Revision 

3, are documented in the topical report 

final safety evaluation report issued on 

October 29, 2019 (ADAMS Accession 

No. ML19297G520). The approved 

version TR–0915–17565–NP–A, 

Revision 4 (ADAMS Accession No. 

ML20057G132) is discussed in the DCA 

safety evaluation report Section 12.2, 

‘‘Radiation Sources,’’ Section 12.3, 

‘‘Radiation Protection Design Features,’’ 

Section 3.11 ‘‘Environmental 

Qualification of Mechanical and 

Electrical Equipment,’’ and Section 

15.0.3, ‘‘Radiological Consequences of 

Design Basis Accidents.’’ The NRC 

found the accident source terms 

acceptable for the purposes described in 

each of the above safety evaluation 

report sections. 

G. Boron Redistribution During Passive 

Cooling Modes 

The NRC evaluated the effects of 

boron volatility and redistribution 

during long term passive cooling. 

During this mode of operation, boronfree 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-ofcycle return to power is discussed in 

Section IV.D, ‘‘Exemption to General 

Design Criterion 27, ‘Combined 

Reactivity Control Systems Capability,’ ’’ 

of this document. In addition, Chapter 

19, Section 19.1.4.6.4, ‘‘Success Criteria, 

Accident Sequences, and Systems 

Analyses,’’ of the safety evaluation 

report concludes that an operator error 

during recovery of the module from an 

uneven boron distribution scenario is 

unlikely to lead to core damage and is 

not a significant risk contributor. 

H. Exemptions 

NuScale Power submitted a total of 17 

requests for exemptions from the 

following regulations, including those 

discussed as part of the significant 

technical issues mentioned previously 

(see Table 1.14–1, ‘‘NuScale Design 

Certification Exemptions,’’ in Chapter 1 

of the final safety evaluation report 

(ADAMS Accession No. 

ML20204A986)): 

1. §§ 50.46a and 50.34(f)(2)(vi) (Reactor 

Coolant System Venting) 

2. § 50.44 (Combustible Gas Control) 

3. § 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. § 50.54(m) (Control Room Staffing) 

(Alternative to meet the regulation) 

7. Appendix A to 10 CFR part 50, GDC 

52, ‘‘Capability for Containment 

Leakage Rate Testing’’ 

8. Appendix A to 10 CFR part 50, GDC 

40, ‘‘Testing of Containment Heat 

Removal System’’ 

9. Appendix A to 10 CFR part 50, GDC 

55, ‘‘Reactor Coolant Pressure 

Boundary Penetrating 

Containment,’’ GDC 56, ‘‘Primary 

Containment Isolation,’’ and GDC 

57, ‘‘Closed Systems Isolation 

Valves’’ (Containment Isolation) 

10. Appendix K to 10 CFR part 50 

(Emergency Core Cooling System 

Evaluation Models) 

11. § 50.34(f)(2)(xx) (Power Supplies for 

Pressurizer Relief Valves, Block 

Valves, and Level Indicators) 

12. § 50.34(f)(2)(xiii) (Pressurizer Heater 

Power Supplies) 

13. § 50.34(f)(2)(xiv)(E) (Containment 

Evacuation System Isolation) 

14. § 50.46 (Fuel Rod Cladding Material) 

15. Appendix A to 10 CFR part 50, GDC 

27, ‘‘Combined Reactivity Control 

Systems Capability’’ 

16. § 50.34(f)(2)(viii) (Post-Accident 

Sampling) 

17. Appendix A to 10 CFR part 50, GDC 

19, ‘‘Control Room’’ 

NRC’s safety evaluation report for 

Chapter 1, ‘‘Introduction and General 

Discussion’’ Section 1.14, ‘‘Index of 

Exemptions,’’ lists these exemption 

requests with the corresponding 

sections of the safety evaluation reports 

where these exemption requests have 

been evaluated. The NRC granted each 

exemption request. 

V. Discussion 

Final Safety Evaluation Report 

NuScale Power submitted the final 

revision of the NuScale DCA, Revision 

5, in July 2020 (ADAMS Accession No. 

ML20225A071). In August 2020, the 

NRC issued a final safety evaluation 

report (ADAMS Accession No. 

ML20023A318) after the Advisory 

Committee on Reactor Safeguards 

(ACRS) performed its final independent 

review and issued its letter to the 

Commission in July 2020 on its findings 


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and recommendations (ADAMS 

Accession No. ML20211M386). The 

final safety evaluation report is a 

collection of reports written by the NRC 

documenting the safety findings from its 

review of the standard design 

application, and it reflects all changes 

resulting from interactions with the 

ACRS as well as changes in the final 

version of the DCA. The final safety 

evaluation report reflects that NuScale 

Power has resolved all technical and 

safety issues with the exception of the 

three issues discussed previously. The 

final safety evaluation report describes 

the portions of the design that are not 

receiving finality in this rule and, 

therefore, will not be part of the 

certified design. The final safety 

evaluation report includes an index of 

all NRC requests for additional 

information, a chronology of all 

documents related to the NuScale DCA 

review, and summaries of public 

meetings and audits. 

NuScale Design Certification Proposed 

Rule 

The following discussion describes 

the purpose and key aspects of each 

section of this NuScale design 

certification proposed rule. All section 

and paragraph references are to the 

provisions being added as appendix G 

to 10 CFR part 52, unless otherwise 

noted. The NRC has modeled this 

NuScale design certification proposed 

rule on existing design certification 

rules, with certain modifications where 

necessary to account for differences in 

the design documentation, design 

features, and environmental assessment 

(including severe accident mitigation 

design alternatives). As a result, design 

certification rules are standardized to 

the extent practical. 

A. Introduction (Section I) 

The purpose of Section I of appendix 

G to 10 CFR part 52 is to identify the 

standard design that would be approved 

by this design certification proposed 

rule and the applicant for certification 

of the standard design. Identification of 

the design certification applicant is 

necessary to implement appendix G to 

10 CFR part 52 for two reasons. First, 

the implementation of § 52.63(c) 

depends on whether an applicant for a 

COL contracts with the design 

certification applicant to obtain the 

generic DCD and supporting design 

information. If the COL applicant does 

not use the design certification 

applicant to provide the design 

information and instead uses an 

alternate nuclear plant vendor, then the 

COL applicant must meet the 

requirements in § 52.73. Second, 

paragraph X.A.1 would require that the 

identified design certification applicant 

maintain the generic DCD throughout 

the time that appendix G to 10 CFR part 

52 may be referenced. 

B. Definitions (Section II) 

The purpose of Section II of appendix 

G to 10 CFR part 52 is to define specific 

terminology with respect to this design 

certification proposed rule. During 

development of the first two design 

certification rules, the NRC decided that 

there would be both generic DCDs 

maintained by the NRC and the design 

certification applicant, as well as 

individual plant-specific DCDs 

maintained by each applicant or 

licensee that references a 10 CFR part 52 

appendix. This distinction is necessary 

in order to specify the relevant plantspecific 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 plantspecific DCD must either include or 

incorporate by reference the information 

in the generic DCD. The COL licensee 

will be required to maintain the plantspecific DCD, updating it as necessary to 

reflect the generic changes to the DCD 

that the NRC may adopt through 

rulemaking, plant-specific departures 

from the generic DCD that the NRC 

imposes on the licensee by order, and 

any plant-specific departures that the 

licensee chooses to make in accordance 

with the relevant processes in Section 

VIII of appendix G to 10 CFR part 52. 

A COL applicant may also have to 

include considerations for multi-module 

facilities in the plant-specific DCD that 

were not previously evaluated as part of 

the design certification rule, depending 

on the contents of the application. 

Therefore, the plant-specific DCD 

functions like an updated final safety 

analysis report because it would provide 

the most complete and accurate 

information on a plant’s design basis for 

that part of the plant that would be 

within the scope of appendix G to 10 

CFR part 52. 

The NRC is treating the technical 

specifications in Chapter 16, ‘‘Technical 

Specifications,’’ of the generic DCD as a 

special category of information and 

designating them as generic technical 

specifications in order to facilitate the 

special treatment of this information 

under appendix G to 10 CFR part 52. A 

COL applicant must submit plantspecific 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 § 52.79 will 

consist of the plant-specific DCD, the 

site-specific final safety analysis report, 

and the plant-specific technical 

specifications. 

The terms Tier 1, Tier 2, and COL 

items (license information) are defined 

in appendix G to 10 CFR part 52 

because these concepts were not 

envisioned when 10 CFR part 52 was 

developed. The design certification 

applicants and the NRC use these terms 

in implementing a two-tiered rule 

structure (the DCD is divided into Tier 

1 and Tier 2 to support the rule 

structure) that was proposed by 

representatives of the nuclear industry 

after publication of 10 CFR part 52. The 

Commission approved the use of the 

two-tiered rule structure in its staff 

requirements memorandum, dated 

February 15, 1991, on SECY–90–377, 

‘‘Requirements for Design Certification 

under 10 CFR part 52,’’ dated November 

8, 1990 (ADAMS Accession No. 

ML003707892). 

Tier 1 information means the portion 

of the design-related information 

contained in the generic DCD that is 

approved and certified by this 

appendix. Tier 2 information means the 

portion of the design-related 

information contained in the generic 

DCD that is approved but not certified 

by this appendix. The change process 

for Tier 2 information is similar, but not 

identical to, the change process set forth 

in § 50.59. The regulations in § 50.59 

describe when a licensee may make 

changes to a plant as described in its 

final safety analysis report without a 


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license amendment. Because of some 

differences in how the change control 

requirements are structured in the 

design certification rules, certain 

definitions contained in § 50.59 are not 

applicable to 10 CFR part 52 and are not 

being included in this proposed rule. 

The NRC is including a definition for 

‘‘Departure from a method of 

evaluation’’ in paragraph II.F of 

appendix G to 10 CFR part 52, so that 

the eight criteria in paragraph VIII.B.5.b 

will be implemented for new reactors as 

intended. 

C. Scope and Contents (Section III) 

The purpose of Section III of 

appendix G to 10 CFR part 52 is to 

describe and define the scope and 

content of this design certification, 

explain how to obtain a copy of the 

generic DCD, identify requirements for 

incorporation by reference of the design 

certification rule, and set forth how 

documentation discrepancies or 

inconsistencies are to be resolved. 

Paragraph III.A is the required 

statement of the Office of the Federal 

Register for approval of the 

incorporation by reference of the 

NuScale DCD, Revision 5. In addition, 

this paragraph provides the information 

on how to obtain a copy of the DCD. 

Unlike previous design certifications, 

the documents submitted to the NRC by 

NuScale Power did not use the title 

‘‘Design Control Document;’’ they used 

the title ‘‘Design Certification 

Application’’ instead. 

Paragraph III.B is the requirement for 

COL applicants and licensees 

referencing the NuScale DCD. The legal 

effect of incorporation by reference is 

that the incorporated material has the 

same legal status as if it were published 

in the Code of Federal Regulations. This 

material, like any other properly issued 

regulation, has the force and effect of 

law. Tier 1 and Tier 2 information 

(including the technical and topical 

reports referenced in the DCD Tier 2, 

Chapter 1) and generic technical 

specifications have been combined into 

a single document called the generic 

DCD in order to effectively control this 

information and facilitate its 

incorporation by reference into the rule. 

In addition, paragraph III.B clarifies that 

the conceptual design information and 

NuScale Power’s evaluation of severe 

accident mitigation design alternatives 

are not considered to be part of 

appendix G to 10 CFR part 52. As 

provided by § 52.47(a)(24), these 

conceptual designs are not part of 

appendix G to 10 CFR part 52 and, 

therefore, are not applicable to an 

application that references appendix G 

to 10 CFR part 52. Therefore, an 

applicant would not be required to 

conform to the conceptual design 

information that was provided by the 

design certification applicant. The 

conceptual design information, which 

consists of site-specific design features, 

was required to facilitate the design 

certification review. Similarly, the 

severe accident mitigation design 

alternatives were required to facilitate 

the environmental assessment. 

Paragraphs III.C and III.D set forth the 

manner by which potential conflicts are 

to be resolved and identify the 

controlling document. Paragraph III.C 

establishes the Tier 1 description in the 

DCD as controlling in the event of an 

inconsistency between the Tier 1 and 

Tier 2 information in the DCD. 

Paragraph III.D establishes the generic 

DCD as the controlling document in the 

event of an inconsistency between the 

DCD and the final safety evaluation 

report for the certified standard design. 

Paragraph III.E makes it clear that 

design activities outside the scope of the 

design certification may be performed 

using actual site characteristics. This 

provision applies to site-specific 

portions of the plant, such as the 

administration building. 

D. Additional Requirements and 

Restrictions (Section IV) 

Section IV of appendix G to 10 CFR 

part 52 sets forth additional 

requirements and restrictions imposed 

upon an applicant who references 

appendix G to 10 CFR part 52. 

Paragraph IV.A sets forth the 

information requirements for COL 

applicants and distinguishes between 

information and documents that must 

be included in the application or the 

DCD and those which may be 

incorporated by reference. Any 

incorporation by reference in the 

application should be clear and should 

specify the title, date, edition or version 

of a document, the page number(s), and 

table(s) containing the relevant 

information to be incorporated. The 

legal effect of such an incorporation by 

reference into the application is that 

appendix G to 10 CFR part 52 would be 

legally binding on the applicant or 

licensee. 

In paragraph IV.B the NRC reserves 

the right to determine how appendix G 

to 10 CFR part 52 may be referenced 

under 10 CFR part 50. This 

determination may occur in the context 

of a subsequent rulemaking modifying 

10 CFR part 52 or this design 

certification rule, or on a case-by-case 

basis in the context of a specific 

application for a 10 CFR part 50 

construction permit or operating 

license. This provision is necessary 

because the previous design 

certification rules were not 

implemented in the manner that was 

originally envisioned at the time that 10 

CFR part 52 was issued. The NRC’s 

concern is with the manner by which 

the inspections, tests, analyses, and 

acceptance criteria (ITAAC) were 

developed and the lack of experience 

with design certifications in a licensing 

proceeding. Therefore, it is appropriate 

that the NRC retain some discretion 

regarding the manner by which 

appendix G to 10 CFR part 52 could be 

referenced in a 10 CFR part 50 licensing 

proceeding. 

E. Applicable Regulations (Section V) 

The purpose of Section V of appendix 

G to 10 CFR part 52 is to specify the 

regulations that were applicable and in 

effect at the time this design 

certification was approved. These 

regulations consist of the technically 

relevant regulations identified in 

paragraph V.A, except for the 

regulations in paragraph V.B that would 

not be applicable to this certified 

design. 

F. Issue Resolution (Section VI) 

The purpose of Section VI of 

appendix G to 10 CFR part 52 is to 

identify the scope of issues that would 

be resolved by the NRC through this 

proposed rule and, therefore, are 

‘‘matters resolved’’ within the meaning 

and intent of § 52.63(a)(5). The section 

is divided into five parts: Paragraph 

VI.A identifies the NRC’s safety findings 

in adopting appendix G to 10 CFR part 

52, paragraph VI.B identifies the scope 

and nature of issues that would be 

resolved by this proposed rule, 

paragraph VI.C identifies issues which 

are not resolved by this proposed rule, 

and paragraph VI.D identifies the issue 

finality restrictions applicable to the 

NRC with respect to appendix G to 10 

CFR part 52. 

Paragraph VI.A describes the nature of 

the NRC’s findings in general terms and 

makes the findings required by § 52.54 

for the NRC’s approval of this design 

certification proposed rule. 

Paragraph VI.B sets forth the scope of 

issues that may not be challenged as a 

matter of right in subsequent 

proceedings. The introductory phrase of 

paragraph VI.B clarifies that issue 

resolution, as described in the 

remainder of the paragraph, extends to 

the delineated NRC proceedings 

referencing appendix G to 10 CFR part 

52. The remainder of paragraph VI.B 

describes the categories of information 

for which there is issue resolution. 

Paragraph VI.C reserves the right of 

the NRC to impose operational 


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1Certain activities ordinarily conducted 

following fuel load and, therefore, considered 

‘‘operational requirements,’’ but which may be 

relied upon to support a Commission finding under 

§ 52.103(g), may themselves be the subject of 

ITAAC to ensure their implementation prior to the 

§ 52.103(g) finding. 

requirements on applicants that 

reference appendix G to 10 CFR part 52. 

This provision reflects the fact that only 

some operational requirements, 

including portions of the generic 

technical specification in Chapter 16 of 

the DCD, were completely or 

comprehensively reviewed by the NRC 

in this design certification proposed 

rule proceeding. The NRC notes that 

operational requirements may be 

imposed on licensees referencing this 

design certification through the 

inclusion of license conditions in the 

license or inclusion of a description of 

the operational requirement in the 

plant-specific final safety analysis 

report.1 The NRC’s choice of the 

regulatory vehicle for imposing the 

operational requirements will depend 

upon, among other things, (1) whether 

the development and/or implementation 

of these requirements must occur prior 

to either the issuance of the COL or the 

Commission finding under § 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. 

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 

§ 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 nonsafeguards 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 

§ 52.85 or an adjudicatory hearing. For 

proceedings where the notice of hearing 

was published before the effective date 

of the final rule, the Commission’s order 

governing access to SUNSI and SGI 

shall be used to govern access to such 

information within the scope of the 

rulemaking. For proceedings in which 

the notice of hearing or opportunity for 

hearing is published after the effective 

date of the final rule, paragraph VI.E 

applies and governs access to SUNSI 

and SGI. 

G. Duration of This Appendix (Section 

VII) 

The purpose of Section VII of 

appendix G to 10 CFR part 52 is, in part, 

to specify the period during which this 

design certification may be referenced 

by an applicant for a COL, under 

§ 52.55, and the period it will remain 

valid when the design certification is 

referenced. For example, if an 

application references this design 

certification during the 15-year period, 

then the design certification would be 

effective until the application is 

withdrawn or the license issued on that 

application expires. The NRC intends 

for appendix G to 10 CFR part 52 to 

remain valid for the life of any COL that 

references the design certification to 

achieve the benefits of standardization 

and licensing stability. This means that 

changes to, or plant-specific departures 

from, information in the plant-specific 

DCD must be made under the change 

processes in Section VIII for the life of 

the plant. 

H. Processes for Changes and Departures 

(Section VIII) 

The purpose of Section VIII of 

appendix G to 10 CFR part 52 is to set 

forth the processes for generic changes 

to, or plant-specific departures 

(including exemptions) from, the DCD. 

The NRC adopted this restrictive change 

process in order to achieve a more stable 

licensing process for applicants and 

licensees that reference design 

certification rules. Section VIII is 

divided into three paragraphs, which 

correspond to Tier 1, Tier 2, and 

operational requirements. 

Generic changes (called 

‘‘modifications’’ in § 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 § 52.63(a)(1). 

Consistent with § 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 § 50.59 

departure or an exemption. Because 

these plant-specific departures will 

result in a DCD that is unique for that 

plant, Section X would require an 

applicant or licensee to maintain a 

plant-specific DCD. For purposes of 

brevity, the following discussion refers 

to the processes for both generic 

changes and plant-specific departures as 

‘‘change processes.’’ Section VIII refers 

to an exemption from one or more 

requirements of this appendix and 

addresses the criteria for granting an 

exemption. The NRC cautions that when 

the exemption involves an underlying 

substantive requirement (i.e., a 

requirement outside this appendix), 

then the applicant or licensee requesting 

the exemption must demonstrate that an 

exemption from the underlying 

applicable requirement meets the 

criteria of §§ 52.7 and 50.12. 

For the NuScale review, the staff 

followed the approach described in 

SECY–17–0075, ‘‘Planned 

Improvements in Design Certification 

Tiered Information Designations,’’ dated 

July 24, 2017 (ADAMS Accession No. 

ML16196A321), to evaluate the 

applicant’s designation of information 

as Tier 1 or Tier 2 information. Unlike 

some of the prior DCAs, this application 

did not contain any Tier 2* information. 

As described in SECY–17–0075, prior 

design certification rules in 10 CFR part 

52, appendices A through E, 

information contained in the DCD was 

divided into three designations: Tier 1, 

Tier 2, and Tier 2*. Tier 1 information 

is the portion of design-related 

information in the generic DCD that the 

Commission approves in the 10 CFR 

part 52 design certification rule 

appendices. To change Tier 1 

information, NRC approval by 

rulemaking or approval of an exemption 

from the certified design rule is 

required. Tier 2 information is also 

approved by the Commission in the 10 

CFR part 52 design certification rule 


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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 § 50.59 and is 

generally referred to as the ‘‘50.59-like’’ 

process. If the criteria in Section VIII are 

met, a licensee can change Tier 2 

information without prior NRC 

approval. 

As mentioned in the previous 

paragraph, the NRC has used a third 

category, Tier 2*, in other design 

certification rules. This third category 

was created to address industry requests 

to minimize the scope of Tier 1 

information and provide greater 

flexibility for making changes. Unlike 

Tier 2 information, all changes to Tier 

2* information require a license 

amendment, but unlike Tier 1 

information, no exemption is required. 

In those rules, Tier 2* information has 

the same safety significance as Tier 1 

information but is part of the Tier 2 

section of the DCD to afford more 

flexibility for licensees to change this 

type of information. 

The applicant did not designate or 

categorize any Tier 2* information in 

the NuScale DCA. The NRC evaluated 

the Tier 2 information to determine 

whether any of that information should 

require NRC approval before it is 

changed. If the NRC had identified any 

such information in Tier 2, then the 

NRC would have requested that the 

applicant revise the application to 

categorize that information as Tier 1 or 

Tier 2*. The NRC did not identify any 

information in Tier 2 that should be 

categorized as Tier 2*. Because neither 

the applicant nor the NRC have 

designated any information in the DCD 

as Tier 2*, that designation and related 

requirements are not being used in this 

design certification rule. 

Tier 1 Information 

Paragraph A of Section VIII describes 

the change process for changes to Tier 

1 information that are accomplished by 

rulemakings that amend the generic 

DCD and are governed by the standards 

in § 52.63(a)(1). A generic change under 

§ 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 § 52.63(a)(2). The NRC 

will give consideration as to whether 

the benefits justify the costs for plants 

that are already licensed or for which an 

application for a permit or license is 

under consideration. 

Departures from Tier 1 may occur in 

two ways: (1) The NRC may order a 

licensee to depart from Tier 1, as 

provided in paragraph VIII.A.3; or (2) an 

applicant or licensee may request an 

exemption from Tier 1, as addressed in 

paragraph VIII.A.4. If the NRC seeks to 

order a licensee to depart from Tier 1, 

paragraph VIII.A.3 would require that 

the NRC find both that the departure is 

necessary for adequate protection or for 

compliance and that special 

circumstances are present. Paragraph 

VIII.A.4 would provide that exemptions 

from Tier 1 requested by an applicant or 

licensee are governed by the 

requirements of §§ 52.63(b)(1) and 

52.98(f), which provide an opportunity 

for a hearing. In addition, the NRC 

would not grant requests for exemptions 

that may result in a significant decrease 

in the level of safety otherwise provided 

by the design. 

Tier 2 Information 

Paragraph B of Section VIII describes 

the change processes for the Tier 2 

information; which have the same 

elements as the Tier 1 change process, 

but some of the standards for plantspecific 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 

§ 52.63(a)(1). A generic change under 

§ 52.63(a)(1) would not be made to a 

certified design while it is in effect 

unless the change: (1) Is necessary for 

compliance with NRC regulations that 

were applicable and in effect at the time 

the certification was issued; (2) is 

necessary to provide adequate 

protection of the public health and 

safety or common defense and security; 

(3) reduces unnecessary regulatory 

burden and maintains protection to 

public health and safety and common 

defense and security; (4) provides the 

detailed design information necessary to 

resolve select design acceptance criteria; 

(5) corrects material errors in the 

certification information; (6) 

substantially increases overall safety, 

reliability, or security of a facility and 

the costs of the change are justified; or 

(7) contributes to increased 

standardization of the certification 

information. 

Departures from Tier 2 would occur 

in four ways: (1) The NRC may order a 

plant-specific departure, as set forth in 

paragraph VIII.B.3; (2) an applicant or 

licensee may request an exemption from 

a Tier 2 requirement as set forth in 

paragraph VIII.B.4; (3) a licensee may 

make a departure without prior NRC 

approval under paragraph VIII.B.5; or 

(4) the licensee may request NRC 

approval for proposed departures which 

do not meet the requirements in 

paragraph VIII.B.5 as provided in 

paragraph VIII.B.5.e. 

Similar to ordered Tier 1 departures 

and generic Tier 2 changes, ordered Tier 

2 departures could not be imposed 

except when necessary, either to bring 

the certification into compliance with 

the NRC’s regulations applicable and in 

effect at the time of approval of the 

design certification or to ensure 

adequate protection of the public health 

and safety or common defense and 

security, as set forth in paragraph 

VIII.B.3. However, unlike Tier 1 

departures, the Commission would not 

have to consider whether the special 

circumstances for the Tier 2 departures 

would outweigh any decrease in safety 

that may result from the reduction in 

standardization caused by the plantspecific order, as required by 

§ 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 § 52.63(a)(4) and (b)(1) 

because it would unnecessarily restrict 

the flexibility of applicants and 

licensees with respect to Tier 2 

information. 

An applicant or licensee would be 

permitted to request an exemption from 

Tier 2 information as set forth in 

paragraph VIII.B.4. The applicant or 

licensee would have to demonstrate that 

the exemption complies with one of the 

special circumstances in regulations 

governing specific exemptions in 

§ 50.12(a). In addition, the NRC would 

not grant requests for exemptions that 

may result in a significant decrease in 

the level of safety otherwise provided by 

the design. However, unlike Tier 1 

changes, the special circumstances for 

the exemption do not have to outweigh 

any decrease in safety that may result 

from the reduction in standardization 

caused by the exemption. If the 

exemption is requested by an applicant 


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for a license, the exemption would be 

subject to litigation in the same manner 

as other issues in the licensing hearing, 

consistent with § 52.63(b)(1). If the 

exemption is requested by a licensee, 

then the exemption would be subject to 

litigation in the same manner as a 

license amendment. 

Paragraph VIII.B.5 would allow an 

applicant or licensee to depart from Tier 

2 information, without prior NRC 

approval, if it does not involve a change 

to, or departure from, Tier 1 

information, technical specification, or 

does not require a license amendment 

under paragraphs VIII.B.5.b or c. The 

technical specifications referred to in 

VIII.B.5.a of this paragraph are the 

technical specifications in Chapter 16 of 

the generic DCD, including bases, for 

departures made prior to the issuance of 

the COL. After the issuance of the COL, 

the plant-specific technical 

specifications would be controlling 

under paragraph VIII.B.5. The 

requirement for a license amendment in 

paragraph VIII.B.5.b would be similar to 

the requirement in § 50.59 and would 

apply to all of the information in Tier 

2 except for the information that 

resolves the severe accident issues or 

the information required by 

§ 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 § 52.47(a)(28). Under 

§ 52.47(a)(28), applicants are required to 

include the information required by 

§ 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 

§ 50.150(a). The applicant or licensee is 

also required to describe in the plantspecific DCD how the modified design 

features and functional capabilities 

continue to meet the assessment 

requirements in § 50.150(a)(1). 

Submittal of this updated information is 

governed by the reporting requirements 

in Section X.B. 

During an ongoing adjudicatory 

proceeding (e.g., for issuance of a COL), 

a party who believes that an applicant 

or licensee has not complied with 

paragraph VIII.B.5 when departing from 

Tier 2 information may petition to admit 

such a contention into the proceeding 

under paragraph VIII.B.5.g. As set forth 

in paragraph VIII.B.5.g, the petition 

would have to comply with the 

requirements of § 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 

§ 2.206. 

Operational Requirements 

The change process for technical 

specifications and other operational 

requirements that were reviewed and 

approved in the design certification rule 

is set forth in Section VIII, paragraph C. 

The key to using the change processes 

described in Section VIII is to determine 

if the proposed change or departure 

would require a change to a design 

feature described in the generic DCD. If 

a design change is required, then the 

appropriate change process in paragraph 

VIII.A or VIII.B would apply. However, 

if a proposed change to the technical 

specifications or other operational 

requirements does not require a change 

to a design feature in the generic DCD, 

then paragraph VIII.C would apply. This 

change process has elements similar to 

the Tier 1 and Tier 2 change processes 

in paragraphs VIII.A and VIII.B, but 

with significantly different change 

standards. Because of the different 

finality status for technical 

specifications and other operational 

requirements, the NRC designated a 

special category of information, 

consisting of the technical specifications 

and other operational requirements, 

with its own change process in 

paragraph VIII.C. The language in 

paragraph VIII.C also distinguishes 

between generic (Chapter 16 of the DCD) 

and plant-specific technical 

specifications to account for the 

different treatment and finality 

consistent with technical specifications 

before and after a license is issued. 

The process in paragraph VIII.C.1 for 

making generic changes to the generic 

technical specifications in Chapter 16 of 

the DCD or other operational 

requirements in the generic DCD would 

be accomplished by rulemaking and 

governed by the backfit standards in 

§ 50.109. The determination of whether 

the generic technical specifications and 

other operational requirements were 

completely reviewed and approved in 

the design certification rule would be 

based upon the extent to which the NRC 

reached a safety conclusion in the final 

safety evaluation report on this matter. 

If a technical specification or 

operational requirement was completely 

reviewed and finalized in the design 

certification rule, then the requirement 

of § 50.109 would apply because a 

position was taken on that safety matter. 

Generic changes made under paragraph 

VIII.C.1 would be applicable to all 

applicants or licensees (refer to 

paragraph VIII.C.2), unless the change is 

irrelevant because of a plant-specific 

departure. 

Some generic technical specifications 

contain values in brackets [ ]. The 

brackets are placeholders indicating that 

the NRC’s review is not complete, and 

represent a requirement that the 

applicant for a COL referencing the 

NuScale design certification rule must 

replace the values in brackets with final 

plant-specific values (refer to guidance 

provided in Regulatory Guide 1.206, 

Revision 1, ‘‘Applications for Nuclear 

Power Plants,’’ dated October 2018 

(ADAMS Accession No. 

ML18131A181)). The values in brackets 

are neither part of the design 

certification rule nor are they binding. 

Therefore, the replacement of bracketed 

values with final plant-specific values 

does not require an exemption from the 

generic technical specifications. 

Plant-specific departures may occur 

by either an order under paragraph 

VIII.C.3 or an applicant’s exemption 

request under paragraph VIII.C.4. The 

basis for determining if the technical 

specification or operational requirement 

was completely reviewed and approved 

for these processes would be the same 

as for paragraph VIII.C.1 previously 

discussed. If the technical specifications 

or operational requirement was 

comprehensively reviewed and 

finalized in the design certification rule, 

then the NRC must demonstrate that 

special circumstances are present before 

ordering a plant-specific departure. If 

not, there would be no restriction on 

plant-specific changes to the technical 

specifications or operational 

requirements, prior to the issuance of a 

license, provided a design change is not 

required. Although the generic technical 

specifications were reviewed and 

approved by the NRC in support of the 

design certification review, the NRC 

intends to consider the lessons learned 

from subsequent operating experience 

during its licensing review of the plantspecific technical specifications. The 

process for petitioning to intervene on a 

technical specification or operational 

requirement contained in paragraph 

VIII.C.5 would be similar to other issues 

in a licensing hearing, except that the 

petitioner must also demonstrate why 

special circumstances are present 

pursuant to § 2.335. 

Paragraph VIII.C.6 states that the 

generic technical specifications would 

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specific technical specifications after 

the issuance of a license that references 

this appendix and the change process. 

After a license is issued, the bases for 

the plant-specific technical specification 

would be controlled by the bases change 

provision set forth in the administrative 

controls section of the plant-specific 

technical specifications. 

I. [RESERVED] (Section IX) 

This section is reserved for future use. 

The matters discussed in this section of 

earlier design certification rules— 

inspections, tests, analyses, and 

acceptance criteria—are now addressed 

in the substantive provisions of 10 CFR 

part 52. Accordingly, there is no need to 

repeat these regulatory provisions in the 

NuScale design certification rule. 

However, this section is being reserved 

to maintain consistent section 

numbering with other design 

certification rules. 

J. Records and Reporting (Section X) 

The purpose of Section X of appendix 

G to 10 CFR part 52 is to set forth the 

requirements that will apply to 

maintaining records of changes to and 

departures from the generic DCD, which 

are to be reflected in the plant-specific 

DCD. Section X also sets forth the 

requirements for submitting reports 

(including updates to the plant-specific 

DCD) to the NRC. This section of 

appendix G to 10 CFR part 52 is similar 

to the requirements for records and 

reports in 10 CFR part 50, except for 

minor differences in information 

collection and reporting requirements. 

Paragraph X.A.1 requires that a 

generic DCD including referenced 

SUNSI and SGI be maintained by the 

applicant for this proposed rule. The 

generic DCD concept was developed, in 

part, to meet the requirements for 

incorporation by reference, including 

public availability of documents 

incorporated by reference. However, the 

SUNSI and SGI could not be included 

in the generic DCD because they are not 

publicly available. Nonetheless, the 

SUNSI and SGI were reviewed by the 

NRC and, as stated in paragraph VI.B.2, 

the NRC would consider the 

information to be resolved within the 

meaning of § 52.63(a)(5). Because this 

information, or its equivalent, is not in 

the generic DCD, it is required to be 

provided by an applicant for a license 

referencing this design certification rule. 

Only the generic DCD is identified and 

incorporated by reference into this rule. 

The generic DCD and the NRC approved 

version of the SUNSI and SGI must be 

maintained by the applicant (NuScale 

Power) for the period of time that 

appendix G to 10 CFR part 52 may be 

referenced. 

Paragraphs X.A.2 and X.A.3 place 

recordkeeping requirements on the 

applicant or licensee that reference this 

design certification so that its plantspecific DCD accurately reflects both 

generic changes to the generic DCD and 

plant-specific departures made under 

Section VIII. The term ‘‘plant-specific’’ 

is used in paragraph X.A.2 and other 

sections of appendix G to 10 CFR part 

52 to distinguish between the generic 

DCD that would be incorporated by 

reference into appendix G to 10 CFR 

part 52, and the plant-specific DCD that 

the COL applicant is required to submit 

under paragraph IV.A. The requirement 

to maintain changes to the generic DCD 

is explicitly stated to ensure that these 

changes are not only reflected in the 

generic DCD, which will be maintained 

by the applicant for the design 

certification, but also in the plantspecific 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 § 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 § 50.150(a) 

throughout the pendency of the 

application and for the term of the 

license and any renewal. This provision 

is consistent with § 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 § 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 

§ 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 § 52.79), but is not part of 

the plant-specific DCD (required by 

paragraph IV.A). Therefore, this 

proposed rule does not require that 

duplicate documentation be maintained 

by an applicant or licensee that 

references this rule because the plantspecific DCD is part of the final safety 

analysis report for the facility. 

Paragraph X.B.1 requires applicants or 

licensees that reference this rule to 

submit reports that describe departures 

from the DCD and include a summary 

of the written evaluations. The 

requirement for the written evaluations 

is set forth in paragraph X.A.3. The 

frequency of the report submittals is set 

forth in paragraph X.B.3. The 

requirement for submitting a summary 

of the evaluations will be similar to the 

requirement in § 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 plantspecific 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 semiannual 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 § 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 

§ 52.103(g)), the frequency of reporting 

will be governed by the requirements in 

paragraph X.B.3.c. 


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VI. Section-by-Section Analysis 

The following paragraphs describe the 

specific changes of this proposed rule: 

Section 52.11, Information collection 

requirements: Office of Management 

and Budget (OMB) approval. 

In § 52.11, this proposed rule would 

add new appendix G to 10 CFR part 52 

to the list of information collection 

requirements in paragraph (b) of this 

section. 

Appendix G to Part 52—Design 

Certification Rule for the NuScale 

Standard Design 

This proposed rule would add 

appendix G to 10 CFR part 52 to 

incorporate the NuScale standard design 

into the NRC’s regulations. Applicants 

intending to construct and operate a 

plant using NuScale may do so by 

referencing the design certification rule. 

VII. Regulatory Flexibility Certification 

Under the Regulatory Flexibility Act 

(5 U.S.C. 605(b)), the NRC certifies that 

this rule, if promulgated, will not have 

a significant economic impact on a 

substantial number of small entities. 

This proposed rule affects only the 

licensing and operation of nuclear 

power plants. The companies that own 

these plants do not fall within the scope 

of the definition of ‘‘small entities’’ set 

forth in the Regulatory Flexibility Act or 

the size standards established by the 

NRC (§ 2.810). 

VIII. Regulatory Analysis 

The NRC has not prepared a 

regulatory analysis for this proposed 

rule. The NRC prepares regulatory 

analyses for rulemakings that establish 

generic regulatory requirements 

applicable to all licensees. Design 

certifications are not generic 

rulemakings in the sense that design 

certifications do not establish standards 

or requirements with which all 

licensees must comply. Rather, design 

certifications are NRC approvals of 

specific nuclear power plant designs by 

rulemaking, which then may be 

voluntarily referenced by applicants for 

combined licenses. Furthermore, design 

certification rules are requested by an 

applicant for a design certification, 

rather than the NRC. Preparation of a 

regulatory analysis in this circumstance 

would not be useful because the design 

to be certified is proposed by the 

applicant rather than the NRC. For these 

reasons, the NRC concludes that 

preparation of a regulatory analysis is 

neither required nor appropriate. 

IX. Backfitting and Issue Finality 

The NRC has determined that this 

proposed rule does not constitute a 

backfit as defined in the backfit rule 

(§ 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 (§ 50.109) 

because there are no operating licenses 

under 10 CFR part 50 referencing this 

design certification proposed rule. 

This initial design certification rule is 

not inconsistent with any applicable 

issue finality provision in 10 CFR part 

52 because it does not impose new or 

changed requirements on existing 

design certification rules in appendices 

A through F to 10 CFR part 52, and no 

combined licenses, construction 

permits, or manufacturing licenses 

issued by the NRC at this time reference 

this design certification proposed rule. 

For these reasons, neither a backfit 

analysis nor a discussion addressing the 

issue finality provisions in 10 CFR part 

52 was prepared for this proposed rule. 

X. Plain Writing 

The Plain Writing Act of 2010 (Pub. 

L. 111–274) requires Federal agencies to 

write documents in a clear, concise, 

well-organized manner that also follows 

other best practices appropriate to the 

subject or field and the intended 

audience. The NRC has written this 

document to be consistent with the 

Plain Writing Act as well as the 

Presidential Memorandum, ‘‘Plain 

Language in Government Writing,’’ 

published June 10, 1998 (63 FR 31883). 

The NRC requests comment on the 

proposed rule with respect to clarity 

and effectiveness of the language used. 

XI. Environmental Assessment and 

Finding of No Significant Impact 

The NRC conducted an environmental 

assessment (ADAMS Accession No. 

ML19303C179) and has determined 

under the National Environmental 

Policy Act of 1969, as amended (NEPA), 

and the NRC’s regulations in subpart A 

of 10 CFR part 51, that this proposed 

rule, if adopted, would not be a major 

Federal action significantly affecting the 

quality of the human environment and, 

therefore, an environmental impact 

statement is not required. The NRC’s 

generic determination in this regard is 

reflected in § 51.32(b)(1). The 

Commission has determined in § 51.32 

that there is no significant 

environmental impact associated with 

the issuance of a standard design 

certification or a design certification 

amendment, as applicable. Comments 

on the environmental assessment will 

be limited to the consideration of severe 

accident mitigation design alternatives 

as required by § 51.30(d). 

The basis for the NRC’s categorical 

exclusion in this regard, as discussed in 

the 2007 final rule amending 10 CFR 

parts 51 and 52 (72 FR 49352; August 

28, 2007), is based upon consideration 

that a design certification rule does not 

authorize the siting, construction, or 

operation of a facility referencing any 

particular design; it only codifies the 

NuScale design in a rule. The NRC will 

evaluate the environmental impacts and 

issue an environmental impact 

statement as appropriate under NEPA as 

part of the application for the 

construction and operation of a facility 

referencing any particular DC rule. 

Consistent with § 51.30(d) and 

§ 51.32(b), the NRC has prepared an 

environmental assessment (ADAMS 

Accession No. ML19303C179) for the 

NuScale design addressing various 

design alternatives to prevent and 

mitigate severe accidents. The 

environmental assessment is based, in 

part, upon the NRC’s review of NuScale 

Power’s evaluation of various design 

alternatives to prevent and mitigate 

severe accidents in Revision 5 of the 

DCA Part 3, ‘‘Application Applicant’s 

Environmental Report—Standard 

Design Certification’’ (ADAMS 

Accession No. ML20224A512). Based on 

a review of NuScale Power’s evaluation, 

the NRC concludes that: (1) NuScale 

Power identified a reasonably complete 

set of potential design alternatives to 

prevent and mitigate severe accidents 

for the NuScale design and (2) none of 

the potential design alternatives 

appropriate at the design certification 

stage are justified on the basis of costbenefit considerations. These issues are 

considered resolved for the NuScale 

design. 

Based on its own independent 

evaluation, the NRC concluded that 

none of the possible candidate design 

alternatives appropriate at this design 

certification stage are potentially cost 

beneficial for NuScale for accident 

events. This independent evaluation 

was based on reasonable treatment of 

costs, benefits, and sensitivities. The 

NRC’s conclusion is applicable for sites 

with site characteristics that fall within 

those site parameters specified in the 

NuScale environmental report. The NRC 

concludes that NuScale Power has 

adequately identified areas appropriate 

at this design certification stage where 

risk potentially could be reduced in a 

cost beneficial manner and that NuScale 

Power has adequately assessed whether 

the implementation of the identified 

potential severe accident mitigation 

design alternatives (SAMDAs) or 

candidate design alternatives would be 

cost beneficial for the given site 

parameters. Site-specific SAMDAs, 


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multi-unit aspects, procedural and 

training SAMDAs, and the reactor 

building crane design would need to be 

assessed when a specific site is 

proposed for constructing and operating 

a NuScale power plant. 

The determination of this 

environmental assessment is that there 

will be no significant offsite impact to 

the public from this action. The 

environmental assessment is available 

as indicated under Section XV of this 

proposed rule. 

XII. Paperwork Reduction Act 

This proposed rule contains new or 

amended collections of information 

subject to the Paperwork Reduction Act 

of 1995 (44 U.S.C. 3501 et seq). This 

proposed rule has been submitted to the 

OMB for review and approval of the 

information collections. 

Type of submission: Revision. 

The title of the information collection: 

Appendix G to 10 CFR part 52 Design 

Certification Rule for NuScale. 

The form number if applicable: NA. 

How often the collection is required or 

requested: On occasion 

Who will be required or asked to 

respond: Applicant for a combined 

license, construction permit, or a design 

certification amendment. 

An estimate of the number of annual 

responses: 5 (2 annual responses and 3 

recordkeepers). 

The estimated number of annual 

respondents: 3. 

An estimate of the total number of 

hours needed annually to comply with 

the information collection requirement 

or request: 389 hours (346 reporting 

hours + 43 recordkeeping hours). 

Abstract: The NRC is proposing to 

amend its regulations to certify the 

NuScale standard design. This action is 

necessary so that applicants or licensees 

intending to construct and operate an 

NuScale standard design may do so by 

referencing this design certification rule. 

The applicant for certification of the 

NuScale standard design is NuScale 

Power, LLC. 

The NRC is seeking public comment 

on the potential impact of the 

information collection contained in this 

proposed rule and on the following 

issues: 

(1) Is the proposed information 

collection necessary for the proper 

performance of the functions of the 

NRC, including whether the information 

will have practical utility? 

(2) Is the estimate of the burden of the 

proposed information collection 

accurate? 

(3) Is there a way to enhance the 

quality, utility, and clarity of the 

information to be collected? 

(4) How can the burden of the 

proposed information collection on 

respondents be minimized, including 

the use of automated collection 

techniques or other forms of information 

technology? 

A copy of the OMB clearance package 

is available in ADAMS under Accession 

No. ML20242A000 or can be obtained 

free of charge by contacting the NRC’s 

Public Document Room reference staff 

at 1–800–397–4209, at 301–415–4737, 

or by email to PDR.resource@nrc.gov. 

You may obtain information and 

comment submissions related to the 

OMB clearance package by searching on 

https://www.regulations.gov under 

Docket ID NRC–2017–0029. 

You may submit comments on any 

aspect of these proposed information 

collection(s), including suggestions for 

reducing the burden and on the above 

issues, by the following methods: 

• Federal Rulemaking Website: Go to 

https://www.regulations.gov and search 

for Docket ID NRC–2017–0029. 

• Mail comments to: FOIA, Library, 

and Information Collections Branch, 

Office of the Chief Information Officer, 

Mail Stop: T6–A10M, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001 or to the OMB reviewer 

at: OMB Office of Information and 

Regulatory Affairs (3150–0151), Attn: 

Desk Officer for the Nuclear Regulatory 

Commission, 725 17th Street NW, 

Washington, DC 20503; email: oira_

submission@omb.eop.gov. 

Additionally, this proposed rule 

provides procedures for requesting 

access to proprietary and safeguards 

information for preparation of 

comments on the NuScale design 

certification proposed rule. These 

procedures are guidance for completing 

mandatory information collections 

located in 10 CFR parts 9 and 73 that 

are subject to the Paperwork Reduction 

Act of 1995 (44 U.S.C. 3501 et seq.). 

These information collections were 

approved by OMB under approval 

numbers 3150–0043 and 3150–0002. 

Send comments regarding this 

information collection to the FOIA, 

Library, and Information Collections 

Branch (T6–A10M), U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555 0001, or by email to 

Infocollects.Resource@nrc.gov, and to 

the OMB reviewer at: OMB Office of 

Information and Regulatory Affairs 

(3150–0043 and 3150–0002), Attn: Desk 

Officer for the Nuclear Regulatory 

Commission, 725 17th Street NW, 

Washington, DC 20503; email: oira_

submission@omb.eop.gov. 

Submit comments by August 30, 

2021. Comments received after this date 

will be considered if it is practical to do 

so, but the NRC is able to ensure 

consideration only for comments 

received on or before this date. 

Public Protection Notification 

The NRC may not conduct or sponsor, 

and a person is not required to respond 

to, a collection of information unless the 

document requesting or requiring the 

collection displays a currently valid 

OMB control number. 

XIII. Agreement State Compatibility 

Under the ‘‘Policy Statement on 

Adequacy and Compatibility of 

Agreement States Programs,’’ approved 

by the Commission on June 20, 1997, 

and published in the Federal Register 

(62 FR 46517; September 3, 1997), this 

proposed rule is classified as 

compatibility ‘‘NRC.’’ Compatibility is 

not required for Category ‘‘NRC’’ 

regulations. The NRC program elements 

in this category are those that relate 

directly to areas of regulation reserved 

to the NRC by the Atomic Energy Act or 

the provisions of 10 CFR, and although 

an Agreement State may not adopt 

program elements reserved to the NRC, 

it may wish to inform its licensees of 

certain requirements by a mechanism 

that is consistent with a particular 

State’s administrative procedure laws, 

but does not confer regulatory authority 

on the State. 

XIV. Voluntary Consensus Standards 

The National Technology Transfer 

and Advancement Act of 1995, Public 

Law 104–113, requires that Federal 

agencies use technical standards that are 

developed or adopted by voluntary 

consensus standards bodies unless the 

use of such a standard is inconsistent 

with applicable law or otherwise 

impractical. In this proposed rule, the 

NRC intends to certify the NuScale 

standard design for use in nuclear 

power plant licensing under 10 CFR 

parts 50 or 52. Design certifications are 

not generic rulemakings establishing a 

generally applicable standard with 

which all 10 CFR parts 50 and 52 

nuclear power plant licensees must 

comply. Design certifications are 

Commission approvals of specific 

nuclear power plant designs by 

rulemaking. Furthermore, design 

certifications are initiated by an 

applicant for rulemaking, rather than by 

the NRC. This action does not constitute 

the establishment of a standard that 

contains generally applicable 

requirements. 

XV. Availability of Documents 

The documents identified in the 

following table are available to 


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interested persons through one or more 

of the following methods, as indicated. 

Document ADAMS 

accession No. 

SECY–21–0004, ‘‘Proposed Rule: NuScale Small Modular Reactor Design Certification (RIN 3150–AJ98; NRC–2017–0029)’’ .... ML19353A003 

Staff Requirements Memorandum for SECY–21–0004, ‘‘Proposed Rule: NuScale Small Modular Reactor Design Certification 

(RIN 3150–AJ98; NRC–2017–0029)’’ .............................................................................................................................................. ML21126A153 

NuScale Power, LLC Submittal of the NuScale Standard Plant Design Certification Application (NRC Project No. 0769) (December 2016) .......................................................................................................................................................................................... ML17013A229 

NuScale Power, LLC Submittal of the NuScale Standard Plant Design Certification Application, Revision 5 (July 2020) ............... ML20225A071 

NuScale DCA Final Safety Evaluation Reports (August 2020) .......................................................................................................... ML20023A318 

NuScale Standard Design Certification Application, Part 3, ‘‘Applicant’s Environmental Report—Standard Design Certification,’’ 

Revision 5 (July 2020) ..................................................................................................................................................................... ML20224A512 

Environmental Assessment by the U.S. Nuclear Regulatory Commission Relating to the Certification of the NuScale Standard 

Design .............................................................................................................................................................................................. ML19303C179 

Regulatory History of Design Certification (April 2000) 2 .................................................................................................................... ML003761550 

NuScale Technical and Topical Reports 

ES–0304–1381–NP, Human-System Interface Style Guide, Rev. 4 (December 2019) ..................................................................... ML19338E948 

RP–0215–10815–NP, Concept of Operations, Rev. 3 (May 2019) .................................................................................................... ML19133A293 

RP–0316–17614–NP, Human Factors Engineering Operating Experience Review Results Summary Report, Rev. 0 (December 

2016) ................................................................................................................................................................................................ ML16364A342 

RP–0316–17615–NP, Human Factors Engineering Functional Requirements Analysis and Function Allocation Results Summary 

Report, Rev. 0 (December 2016) .................................................................................................................................................... ML16364A342 

RP–0316–17616–NP, Human Factors Engineering Task Analysis Results Summary Report, Rev. 2 (April 2019) ......................... ML19119A393 

RP–0316–17617–NP, Human Factors Engineering Staffing and Qualifications Results Summary Report, Rev. 0 (December 

2016) ................................................................................................................................................................................................ ML17004A222 

RP–0316–17618–NP, Human Factors Engineering Treatment of Important Human Actions Results Summary Report, Rev. 0 

(December 2016) ............................................................................................................................................................................. ML17004A222 

RP–0316–17619–NP, Human Factors Engineering Human-System Interface Design Results Summary Report, Rev. 2, (April 

2019) ................................................................................................................................................................................................ ML19119A398 

RP–0516–49116–NP, Control Room Staffing Plan Validation Results, Rev. 1 (December 2016) .................................................... ML16364A356 

RP–0914–8534–NP, Human Factors Engineering Program Management Plan, Rev. 5 (April 2019) ............................................... ML19119A342 

RP–0914–8543–NP, Human Factors Verification and Validation Implementation Plan, Rev. 5 (April 2019) .................................... ML19119A372 

RP–0914–8544–NP, Human Factors Engineering Design Implementation Implementation Plan, Rev. 4 (November 2019) ........... ML19331A910 

RP–1018–61289–NP, Human Factors Engineering Verification and Validation Results Summary Report, Rev. 1 (July 2019) ....... ML19212A773 

RP–1215–20253–NP, Control Room Staffing Plan Validation Methodology, Rev. 3 (December 2016) ............................................ ML16364A353 

TR–0116–20781–NP, Fluence Calculation Methodology and Results, Rev. 1 (July 2019) ............................................................... ML19183A485 

TR–0116–20825–NP–A, Applicability of AREVA Fuel Methodology for the NuScale Design, Rev. 1 (February 2018) ................... ML18040B306 

TR–0116–21012–NP–A, NuScale Power Critical Heat Flux Correlations, Rev. 1 (December 2018) ................................................ ML18360A632 

TR–0316–22048–NP, Nuclear Steam Supply System Advanced Sensor Technical Report, Rev. 3 (May 2020) ............................. ML20141M764 

TR–0515–13952–NP–A, Risk Significance Determination, Rev. 0 (October 2016) ........................................................................... ML16284A016 

TR–0516–49084–NP, Containment Response Analysis Methodology Technical Report, Rev. 3 (May 2020) .................................. ML20141L808 

TR–0516–49416–NP–A, Non-Loss-of-Coolant Accident Analysis Methodology, Rev. 3 (July 2020) ................................................ ML20191A281 

TR–0516–49417–NP–A, Evaluation Methodology for Stability Analysis of the NuScale Power Module, Rev. 1 (March 2020) ....... ML20078Q094 

TR–0516–49422–NP–A, Loss-of-Coolant Accident Evaluation Model, Rev. 2 (July 2020) ............................................................... ML20189A644 

TR–0616–48793–NP–A, Nuclear Analysis Codes and Methods Qualification, Rev. 1 (December 2018) ......................................... ML18348B036 

TR–0616–49121–NP, NuScale Instrument Setpoint Methodology Technical Report, Rev. 3 (May 2020) ........................................ ML20141M114 

TR–0716–50350–NP–A, Rod Ejection Accident Methodology, Rev. 1 (June 2020) .......................................................................... ML20168B203 

TR–0716–50351–NP–A, NuScale Applicability of AREVA Method for the Evaluation of Fuel Assembly Structural Response to 

Externally Applied Forces, Rev. 1 (May 2020) ................................................................................................................................ ML20122A248 

TR–0716–50424–NP, Combustible Gas Control, Rev. 1 (March 2019) ............................................................................................. ML19091A232 

TR–0716–50439–NP, NuScale Comprehensive Vibration Assessment Program Analysis Technical Report, Rev. 2 (July 2019) ... ML19212A776 

TR–0815–16497–NP–A, Safety Classification of Passive Nuclear Power Plant Electrical Systems Topical Report, Rev. 1 (February 2018) ....................................................................................................................................................................................... ML18054B607 

TR–0816–49833–NP, Fuel Storage Rack Analysis, Rev. 1 (November 2018) .................................................................................. ML18310A154 

TR–0816–50796–NP, Loss of Large Areas Due to Explosions and Fires Assessment, Rev. 1 (June 2019) ................................... ML19165A294 

TR–0816–50797 (NuScale Nonproprietary), Mitigation Strategies for Loss of All AC Power Event, Rev. 3 (October 2019) ........... ML19302H598 

TR–0816–51127–NP, NuFuel-HTP2TM Fuel and Control Rod Assembly Designs, Rev. 3 (December 2019) .................................. ML19353A719 

TR–0818–61384–NP, Pipe Rupture Hazards Analysis, Rev. 2 (July 2019) ....................................................................................... ML19212A682 

TR–0915–17564–NP–A, Subchannel Analysis Methodology, Rev. 2 (March 2019) .......................................................................... ML19067A256 

TR–0915–17565–NP–A, Accident Source Term Methodology, Rev. 4 (February 2020) ................................................................... ML20057G132 

TR–0916–51299–NP, Long-Term Cooling Methodology, Rev. 3 (May 2020) .................................................................................... ML20141L816 

TR–0916–51502–NP, NuScale Power Module Seismic Analysis, Rev. 2 (April 2019) ...................................................................... ML19093B850 

TR–0917–56119–NP, CNV Ultimate Pressure Integrity, Rev. 1 (June 2019) .................................................................................... ML19158A382 

TR–0918–60894–NP, Comprehensive Vibration Assessment Program Measurement and Inspection Plan Technical Report, 

Rev, 1 (August 2019) ....................................................................................................................................................................... ML19214A248 

TR–1010–859–NP–A, NuScale Topical Report: Quality Assurance Program Description for the NuScale Power Plant, Rev. 5 

(June 2020) ...................................................................................................................................................................................... ML20176A494 

TR–1015–18177–NP, Pressure and Temperature Limits Methodology, Rev. 2 (October 2018) ....................................................... ML18298A304 

TR–1015–18653–NP–A, Design of the Highly Integrated Protection System Platform Topical Report, Rev. 2 (September 2017) ML17256A892 

TR–1016–51669–NP, NuScale Power Module Short-Term Transient Analysis, Rev. 1 (July 2019) ................................................. ML19211D411 

TR–1116–51962–NP, NuScale Containment Leakage Integrity Assurance Technical Report, Rev. 1 (May 2019) .......................... ML19149A298 


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2The 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. 

3Broad SGI requests under these procedures are 

unlikely to meet the standard for need to know. 

Furthermore, NRC redaction of information from 

requested documents before their release may be 

appropriate to comport with this requirement. The 

procedures in this document do not authorize 

unrestricted disclosure or less scrutiny of a 

requester’s need to know than ordinarily would be 

applied in connection with either adjudicatory or 

non-adjudicatory access to SGI. 

Document ADAMS 

accession No. 

TR–1116–52065–NP, Effluent Release (GALE Replacement) Methodology and Results, Rev. 1 (November 2018) ...................... ML18317A364 

The NRC may post materials related 

to this document, including public 

comments, on the Federal Rulemaking 

website at https://www.regulations.gov 

under Docket ID NRC–2017–0029. 

XVI. Procedures for Access to 

Proprietary and Safeguards 

Information for Preparation of 

Comments on the NuScale Design 

Certification Proposed Rule 

This section contains instructions 

regarding how the non-publicly 

available documents related to this rule, 

and specifically those listed in Table 

1.6–1 and 1.6–2 beginning on page 1.6– 

2 of Tier 2 of the DCD, may be accessed 

by interested persons who wish to 

comment on the design certification. 

These documents contain proprietary 

information and safeguards information 

(SGI). Requirements for access to SGI 

are primarily set forth in 10 CFR parts 

2 and 73. This section provides 

information specific to this proposed 

rule; however, nothing in this section is 

intended to conflict with the SGI 

regulations. 

Interested persons who desire access 

to proprietary information on NuScale 

should first request access to that 

information from NuScale Power, LLC, 

the design certification applicant. 

Requests to the applicant must be sent 

to NuScale Power, LLC, at 

RegulatoryAffairs@NuScalePower.com. 

A request for access should be 

submitted to the NRC if the applicant 

does not either grant or deny access by 

the 10-day deadline described in the 

following section. 

One of the non-publicly available 

documents, TR–0416–48929, ‘‘NuScale 

Design of Physical Security Systems,’’ 

contains both proprietary information 

and SGI. If you need access to 

proprietary information in that 

document in order to develop comments 

within the scope of this rule, then your 

request for access should first be 

submitted to NuScale Power, in 

accordance with the previous 

paragraph. By contrast, if you need 

access to the SGI in order to provide 

comments, then your request for access 

to the SGI must be submitted to the NRC 

as described further in this section. 

Therefore, if you need access to both 

proprietary information and SGI in that 

document, then you should request 

access to the information in separate 

requests submitted to both NuScale 

Power and the NRC. 

Submitting a Request to the NRC for 

Access 

Within 10 days after publication of 

this proposed rule, any individual or 

entity who believes access to 

proprietary information or SGI is 

necessary in order to submit comments 

on this proposed rule may request 

access to such information. Requests for 

access to proprietary information or SGI 

submitted more than 10 days after 

publication of this document will not be 

considered absent a showing of good 

cause for the late filing explaining why 

the request could not have been filed 

earlier. 

The requestor shall submit a letter 

requesting permission to access 

proprietary information and/or SGI to 

the Office of the Secretary, U.S. Nuclear 

Regulatory Commission, Attention: 

Rulemakings and Adjudications Staff, 

Washington, DC 20555–0001. The email 

address for the Office of the Secretary is 

Rulemaking.Comments@nrc.gov. The 

requester must send a copy of the 

request to the design certification 

applicant at the same time as the 

original transmission to the NRC using 

the same method of transmission. 

Requests to the applicant must be sent 

to NuScale Power, LLC, at 

RegulatoryAffairs@NuScalePower.com. 

The request must include the 

following information: 

(1) The name of this design 

certification, NuScale Design 

Certification; the rulemaking 

identification number, RIN 3150–AJ98; 

the rulemaking docket number, NRC– 

2017–0029; and the Federal Register 

citation for this rule. 

(2) The name and address of the 

requester. 

(3) The identity of the individual(s) to 

whom access is to be provided, 

including the identity of any expert, 

consultant, or assistant who will aid the 

requestor in evaluating the information. 

(4) If the request is for proprietary 

information, the requester’s need for the 

information in order to prepare 

meaningful comments on the design 

certification must be demonstrated. 

Each of the following areas must be 

addressed with specificity: 

(a) The specific issue or subject matter 

on which the requester wishes to 

comment. 

(b) An explanation why information 

which is publicly available is 

insufficient to provide the basis for 

developing meaningful comment on the 

NuScale design certification proposed 

rule with respect to the issue or subject 

matter described in paragraph 4.a. of 

this section. 

(c) The technical competence 

(demonstrable knowledge, skill, training 

or education) of the requestor to 

effectively utilize the requested 

proprietary information to provide the 

basis for meaningful comment. 

Technical competence may be shown by 

reliance on a qualified expert, 

consultant, or assistant who satisfies 

these criteria. 

(d) A chronology and discussion of 

the requester’s attempts to obtain the 

information from the design 

certification applicant, and the final 

communication from the requester to 

the applicant and the applicant’s 

response, if any was provided, with 

respect to the request for access to 

proprietary information must be 

submitted. 

(5) If the request is for SGI, the request 

must include the following: 

(a) A statement that explains each 

individual’s ‘‘need to know’’ the SGI, as 

required by §§ 73.2 and 73.22(b)(1). 

Consistent with the definition of ‘‘need 

to know’’ as stated in § 73.2, the 

statement must explain: 

(i) Specifically why the requestor 

believes that the information is 

necessary to enable the requestor to 

proffer and/or adjudicate a specific 

contention in this proceeding; 3 and 

(ii) The technical competence 

(demonstrable knowledge, skill, training 

or education) of the requestor to 

effectively utilize the requested SGI to 

provide the basis and specificity for 

meaningful comment. Technical 

competence may be shown by reliance 


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4The requester will be asked to provide his or her 

full name, social security number, date and place 

of birth, telephone number, and email address. 

After providing this information, the requestor 

usually should be able to obtain access to the online 

form within one business day. 

5This fee is subject to change pursuant to DCSA’s 

adjustable billing rates. 

on a qualified expert, consultant, or 

assistant who satisfies these criteria. 

(b) A completed Form SF–85, 

‘‘Questionnaire for Non-Sensitive 

Positions,’’ for each individual who 

would have access to SGI. The 

completed Form SF–85 will be used by 

the Office of Administration to conduct 

the background check required for 

access to SGI, as required by 10 CFR 

part 2, subpart C, and § 73.22(b)(2), to 

determine the requestor’s 

trustworthiness and reliability. For 

security reasons, Form SF–85 can be 

submitted only electronically through 

the Electronic Questionnaires for 

Investigations Processing website, a 

secure website that is owned and 

operated by the Defense 

Counterintelligence and Security 

Agency (DCSA). To obtain online access 

to the form, the requestor should contact 

the NRC’s Office of Administration at 

301–415–3710.4

(c) A completed Form FD–258 

(fingerprint card), signed in original ink, 

and submitted in accordance with 

§ 73.57(d). Copies of Form FD–258 may 

be obtained by sending an email to 

MAILSVC.Resource@nrc.gov or by 

sending a written request to U.S. 

Nuclear Regulatory Commission, Attn: 

Mailroom/Fingerprint Card Request, 

11555 Rockville Pike, Rockville, MD 

20852. The fingerprint card will be used 

to satisfy the requirements of 10 CFR 

part 2, subpart C, § 73.22(b)(1), and 

Section 149 of the Atomic Energy Act of 

1954, as amended, which mandates that 

all persons with access to SGI must be 

fingerprinted for an FBI identification 

and criminal history records check. 

(d) A check or money order in the 

amount of $326.00 5 payable to the U.S. 

Nuclear Regulatory Commission for 

each individual for whom the request 

for access has been submitted; and 

(e) If the requester or any individual 

who will have access to SGI believes 

they belong to one or more of the 

categories of individuals that are exempt 

from the criminal history records check 

and background check requirements, as 

stated in § 73.59, the requester should 

also provide a statement identifying 

which exemption the requester is 

invoking, and explaining the requester’s 

basis for believing that the exemption 

applies. While processing the request, 

the Office of Administration, Personnel 

Security Branch, will make a final 

determination whether the claimed 

exemption applies. Alternatively, the 

requester may contact the Office of 

Administration for an evaluation of 

their exemption status prior to 

submitting their request. Persons who 

are exempt from the background check 

are not required to complete the SF–85 

or Form FD–258; however, all other 

requirements for access to SGI, 

including the need to know, are still 

applicable. 

Note: Copies of documents and 

materials required by paragraphs (5)(b), 

(c), and (d), of this section must be sent 

to the following address: U.S. Nuclear 

Regulatory Commission, ATTN: 

Personnel Security Branch, Mail Stop 

TWFN–07D04M, 11555 Rockville Pike, 

Rockville, MD 20852. 

These documents and materials 

should not be included with the request 

letter to the Office of the Secretary, but 

the request letter should state that the 

forms and fees have been submitted as 

required. 

To avoid delays in processing 

requests for access to SGI, all forms 

should be reviewed for completeness 

and accuracy (including legibility) 

before submitting them to the NRC. The 

NRC will return incomplete or illegible 

packages to the sender without 

processing. 

Based on an evaluation of the 

information submitted under paragraphs 

(4) or (5) of this section, as applicable, 

the NRC will determine within 10 days 

of receipt of the request whether the 

requester has established a legitimate 

need for access to proprietary 

information or need to know the SGI 

requested. 

Determination of Legitimate Need for 

Access 

For proprietary information access 

requests, if the NRC determines that the 

requester has established a legitimate 

need for access to proprietary 

information, the NRC will notify the 

requester in writing that access to 

proprietary information has been 

granted. The written notification will 

contain instructions on how the 

requestor may obtain copies of the 

requested documents, and any other 

conditions that may apply to access to 

those documents. These conditions may 

include, but are not limited to, the 

signing of a Non-Disclosure Agreement 

or Affidavit by each individual who will 

be granted access. 

For requests for access to SGI, if the 

NRC determines that the requester has 

established a need to know the SGI, the 

NRC’s Office of Administration will 

then determine, based upon completion 

of the background check, whether the 

proposed recipient is trustworthy and 

reliable, as required for access to SGI by 

§ 73.22(b). If the NRC’s Office of 

Administration determines that the 

individual or individuals are 

trustworthy and reliable, the NRC will 

promptly notify the requester in writing. 

The notification will provide the names 

of approved individuals as well as the 

conditions under which the SGI will be 

provided. Those conditions may 

include, but are not limited to, the 

signing of a Non-Disclosure Agreement 

or Affidavit by each individual who will 

be granted access to SGI. 

Release and Storage of SGI 

Prior to providing SGI to the 

requester, the NRC will conduct (as 

necessary) an inspection to confirm that 

the recipient’s information protection 

system is sufficient to satisfy the 

requirements of § 73.22. Alternatively, 

recipients may opt to view SGI at an 

approved SGI storage location rather 

than establish their own SGI protection 

program to meet SGI protection 

requirements. 

Filing of Comments on the NuScale 

Design Certification Proposed Rule 

Based on Non-Public Information 

Any comments in this rulemaking 

proceeding that are based upon the 

information received as a result of the 

request made for proprietary or SGI 

information must be filed by the 

requester no later than 25 days after 

receipt of (or access to) that information, 

or the close of the public comment 

period, whichever is later. The 

commenter must comply with all NRC 

requirements regarding the submission 

of proprietary information and SGI to 

the NRC when submitting comments to 

the NRC (including marking and 

transmission requirements). 

Review of Denials of Access 

If the request for access to proprietary 

information or SGI is denied by the 

NRC, either after a determination on 

requisite need or after a determination 

on trustworthiness and reliability, the 

NRC shall promptly notify the requester 

in writing, briefly stating the reason or 

reasons for the denial. 

Before the Office of Administration 

makes a final adverse determination 

regarding the trustworthiness and 

reliability of the proposed recipient(s) 

for access to SGI, the Office of 

Administration, in accordance with 

§ 2.336(f)(1)(iii), must provide the 

proposed recipient(s) any records that 

were considered in the trustworthiness 

and reliability determination, including 

those required to be provided under 

§ 73.57(e)(1), so that the proposed 


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6State-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. 

recipient(s) have an opportunity to 

correct or explain the record. 

The requestor may challenge the 

NRC’s adverse determination with 

respect to access to proprietary 

information or with respect to need to 

know for SGI by filing a challenge 

within 5 days of receipt of that 

determination with the NRC’s Executive 

Director for Operations under § 9.29(d). 

The requestor may challenge the 

Office of Administration’s final adverse 

determination with respect to 

trustworthiness and reliability for access 

to SGI by filing a request for review in 

accordance with § 2.336(f)(1)(iv). 

XVII. Incorporation by Reference— 

Reasonable Availability to Interested 

Parties 

The NRC proposes to incorporate by 

reference the NuScale DCA, Revision 5. 

As described in the ‘‘Discussion’’ 

sections of this document, the generic 

DCD includes Tier 1 and Tier 2 

information (including the technical 

and topical reports referenced in 

Chapter 1) and generic technical 

specifications in order to effectively 

control this information and facilitate its 

incorporation by reference into the rule. 

NuScale Power submitted Revision 5 of 

the DCA to the NRC in July 2020. 

The NRC is required by law to obtain 

approval for incorporation by reference 

from the Office of the Federal Register 

(OFR). The OFR’s requirements for 

incorporation by reference are set forth 

in 1 CFR part 51. The OFR regulations 

require an agency to include in a 

proposed rule a discussion of the ways 

that the materials the agency 

incorporates by reference are reasonably 

available to interested parties or how it 

worked to make those materials 

reasonably available to interested 

parties. The discussion in this section 

complies with the requirement for a 

proposed rule as set forth in 1 CFR 

51.5(a)(1). 

The NRC considers ‘‘interested 

parties’’ to include all potential NRC 

stakeholders, not only the individuals 

and entities regulated or otherwise 

subject to the NRC’s regulatory 

oversight. These NRC stakeholders are 

not a homogenous group but vary with 

respect to the considerations for 

determining reasonable availability. 

Therefore, the NRC distinguishes 

between different classes of interested 

parties for the purposes of determining 

whether the material is ‘‘reasonably 

available.’’ The NRC considers the 

following to be classes of interested 

parties in NRC rulemakings with regard 

to the material to be incorporated by 

reference: 

• Individuals and small entities 

regulated or otherwise subject to the 

NRC’s regulatory oversight (this class 

also includes applicants and potential 

applicants or licenses and other NRC 

regulatory approvals) and who are 

subject to the material to be 

incorporated by reference by 

rulemaking. In this context, ‘‘small 

entities’’ has the same meaning as a 

‘‘small entity’’ under § 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 § 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 § 2.315(c)). 

• Federally-recognized and Staterecognized 6 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: 

Library.Resource@nrc.gov. In addition, 

as described in Section XV of this 

proposed rule, documents related to this 

proposed rule are available online in the 

NRC’s ADAMS Public Documents 

collection at https://www.nrc.gov/ 

reading-rm/adams.html. 

The NRC concludes that the materials 

the NRC is incorporating by reference in 

this proposed rule are reasonably 

available to all interested parties 

because the materials are available in 

multiple ways and in a manner 

consistent with their interest in the 

materials. 

List of Subjects in 10 CFR Part 52 

Administrative practice and 

procedure, Antitrust, Combined license, 

Early site permit, Emergency planning, 

Fees, Incorporation by reference, 

Inspection, Issue finality, Limited work 

authorization, Nuclear power plants and 

reactors, Probabilistic risk assessment, 

Prototype, Reactor siting criteria, 

Redress of site, Penalties, Reporting and 

recordkeeping requirements, Standard 

design, Standard design certification. 

For the reasons set out in the 

preamble and under the authority of the 

Atomic Energy Act of 1954, as amended; 

the Energy Reorganization Act of 1974, 

as amended; the Nuclear Waste Policy 

Act of 1982, as amended; and 5 U.S.C. 

552 and 553, the NRC proposes the 

following amendments to 10 CFR part 

52: 

PART 52—LICENSES, 

CERTIFICATIONS, AND APPROVALS 

FOR NUCLEAR POWER PLANTS 

■ 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. 

§ 52.11 [Amended] 

■ 2. In § 52.11(b), add ‘‘G,’’ in 

alphabetical order to the list of 

appendices. 

■ 3. Add Appendix G to part 52 to read 

as follows: 

Appendix G to Part 52—Design 

Certification Rule for NuScale 

I. Introduction 

Appendix G constitutes the standard 

design certification for NuScale, in 

accordance with 10 CFR part 52, subpart B. 

The applicant for the standard design 

certification of NuScale is NuScale Power, 

LLC. 

II. Definitions 

A. Generic design control document 

(generic DCD) means the document 

containing the Tier 1 and Tier 2 information 

(including the technical and topical reports 

referenced in Chapter 1) and generic 

technical specifications that is incorporated 

by reference into this appendix. 

B. Generic technical specifications (generic 

TS) means the information required by 10 

CFR 50.36 and 50.36a for the portion of the 

plant that is within the scope of this 

appendix. 

C. Plant-specific DCD means that portion of 

the combined license (COL) final safety 

analysis report (FSAR) that sets forth both the 

generic DCD information and any plantspecific changes to generic DCD information. 


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D. Tier 1 means the portion of the designrelated 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 designrelated information contained in the generic 

DCD that is approved but not certified by this 

appendix (Tier 2 information). Compliance 

with Tier 2 is required, but generic changes 

to and plant-specific departures from Tier 2 

are governed by Section VIII of this 

appendix. Compliance with Tier 2 provides 

a sufficient, but not the only acceptable, 

method for complying with Tier 1. 

Compliance methods differing from Tier 2 

must satisfy the change process in Section 

VIII of this appendix G. Regardless of these 

differences, an applicant or licensee must 

meet the requirement in paragraph III.B of 

this appendix to reference Tier 2 when 

referencing Tier 1. Tier 2 information 

includes: 

1. Information required by § 52.47(a) and 

(c), with the exception of generic TS and 

conceptual design information; 

2. Supporting information on the 

inspections, tests, and analyses that will be 

performed to demonstrate that the acceptance 

criteria in the ITAAC have been met; and 

3. COL action items (COL license 

information) identify certain matters that 

must be addressed in the site-specific portion 

of the FSAR by an applicant who references 

this appendix. These items constitute 

information requirements but are not the 

only acceptable set of information in the 

FSAR. An applicant may depart from or omit 

these items, provided that the departure or 

omission is identified and justified in the 

FSAR. After issuance of a construction 

permit or COL, these items are not 

requirements for the licensee unless such 

items are restated in the FSAR. 

F. Departure from a method of evaluation 

described in the plant-specific DCD used in 

establishing the design bases or in the safety 

analyses means: 

1. Changing any of the elements of the 

method described in the plant-specific DCD 

unless the results of the analysis are 

conservative or essentially the same; or 

2. Changing from a method described in 

the plant-specific DCD to another method 

unless that method has been approved by the 

NRC for the intended application. 

G. All other terms in this appendix have 

the meaning set out in 10 CFR 50.2, 10 CFR 

52.1, or Section 11 of the Atomic Energy Act 

of 1954, as amended, as applicable. 

III. Scope and Contents 

A. Incorporation by reference approval. 

NuScale standard design (hereafter referred 

as NuScale) material is approved for 

incorporation by reference by the Director of 

the Office of the Federal Register under 5 

U.S.C. 552(a) and 1 CFR part 51, 

‘‘Incorporation by Reference.’’ You may 

obtain copies of the generic DCD from 

NuScale Power, LLC, 6650 SW Redwood 

Lane, Suite 210, Portland, Oregon 97224. You 

can view the generic DCD online in the NRC 

Library at https://www.nrc.gov/reading-rm/ 

adams.html. In ADAMS, search under 

ADAMS Accession No. ML20225A071. If you 

do not have access to ADAMS or if you have 

problems accessing documents located in 

ADAMS, contact the NRC’s Public Document 

Room (PDR) reference staff at 1–800–397– 

4209, 301–415–3747, or by email at 

PDR.Resource@nrc.gov. Copies of the 

NuScale materials are available in the 

ADAMS Public Documents collection. All 

approved material is available for inspection 

at the National Archives and Records 

Administration (NARA). For information on 

the availability of this material at NARA, 

email at fedreg.legal@nara.gov or go to 

https://www.archives.gov/federal-register/cfr/ 

ibrlocations.html. 

1. NuScale Standard Plant Design 

Certification Application, Certified Design 

Descriptions and Inspections, Tests, 

Analyses, & Acceptance Criteria (ITAAC), 

Part 2—Tier 1, Revision 5, July 2020. 

2. NuScale Standard Plant Design 

Certification Application, Part 2—Tier 2, 

Revision 5, July 2020, including: 

a. Chapter One, Introduction and General 

Description of the Plant. 

b. Chapter Two, Site Characteristics and 

Site Parameters. 

c. Chapter Three, Design of Structures, 

Systems, Components and Equipment. 

d. Chapter Four, Reactor. 

e. Chapter Five, Reactor Coolant System 

and Connecting Systems. 

f. Chapter Six, Engineered Safety Features. 

g. Chapter Seven, Instrumentation and 

Controls. 

h. Chapter Eight, Electric Power. 

i. Chapter Nine, Auxiliary Systems. 

j. Chapter Ten, Steam and Power 

Conversion System. 

k. Chapter Eleven, Radioactive Waste 

Management. 

l. Chapter Twelve, Radiation Protection. 

m. Chapter Thirteen, Conduct of 

Operations. 

n. Chapter Fourteen, Initial Test Program 

and Inspections, Tests, Analyses, and 

Acceptance Criteria. 

o. Chapter Fifteen, Transient and Accident 

Analyses. 

p. Chapter Sixteen, Technical 

Specifications. 

q. Chapter Seventeen, Quality Assurance 

and Reliability Assurance. 

r. Chapter Eighteen, Human Factors 

Engineering. 

s. Chapter Nineteen, Probabilistic Risk 

Assessment and Severe Accident Evaluation. 

t. Chapter Twenty, Mitigation of BeyondDesign-Basis Events. 

u. Chapter Twenty-One, Multi-Module 

Design Considerations. 

3. DCA Part 4, Volume 1, Revision 5.0, 

Generic Technical Specifications, NuScale 

Nuclear Power Plants, Volume 1: 

Specifications. 

4. DCA Part 4, Volume 2, Revision 5.0, 

Generic Technical Specifications, NuScale 

Nuclear Power Plants, Volume 2: Bases. 

5. ES–0304–1381–NP, Human-System 

Interface Style Guide, December 2019, 

Revision 4, Docket: 52–048. 

6. RP–0215–10815–NP, Concept of 

Operations, May 2019, Revision 3, Docket: 

52–048. 

7. RP–0316–17614–NP, Human Factors 

Engineering Operating Experience Review 

Results Summary Report, 12/07/2016, 

Revision 0, Docket: PROJ0769. 

8. RP–0316–17615–NP, Human Factors 

Engineering Functional Requirements 

Analysis and Function Allocation Results 

Summary Report, 12/2/16, Revision 0, 

Docket: PROJ0769. 

9. RP–0316–17616–NP, Human Factors 

Engineering Task Analysis Results Summary 

Report, April 2019, Revision 2, Docket: 52– 

048. 

10. RP–0316–17617–NP, Human Factors 

Engineering Staffing and Qualifications 

Results Summary Report, 12/02/2016, 

Revision 0, Docket: PROJ0769. 

11. RP–0316–17618–NP, Human Factors 

Engineering Treatment of Important Human 

Actions Results Summary Report, 

12/02/2016, Revision 0, Docket: PROJ0769. 

12. RP–0316–17619–NP, Human Factors 

Engineering Human-System Interface Design 

Results Summary Report, April 2019, 

Revision 2, Docket: 52–048. 

13. RP–0516–49116–NP, Control Room 

Staffing Plan Validation Results, 12/02/2016, 

Revision 1, Docket: PROJ0769. 

14. RP–0914–8534–NP, Human Factors 

Engineering Program Management Plan, 

April 2019, Revision 5, Docket: 52–048. 

15. RP–0914–8543–NP, Human Factors 

Verification and Validation Implementation 

Plan, April 2019, Revision 5, Docket: 52–048. 

16. RP–0914–8544–NP, Human Factors 

Engineering Design Implementation 

Implementation Plan, November 2019, 

Revision 4, Docket: 52–048, NuScale 

Nonproprietary. 

17. RP–1018–61289–NP, Human Factors 

Engineering Verification and Validation 

Results Summary Report, July 2019, Revision 

1, Docket: 52–048. 

18. RP–1215–20253–NP, Control Room 

Staffing Plan Validation Methodology, 

12/02/2016, Revision 3, Docket: PROJ0769. 

19. TR–0116–20781–NP, Fluence 

Calculation Methodology and Results, July 

2019, Revision 1, Docket: 52–048. 

20. TR–0116–20825–NP–A, Applicability 

of AREVA Fuel Methodology for the NuScale 

Design, June 2016, Revision 1, Docket: 

PROJ0769. 

21. TR–0116–21012–NP–A, NuScale Power 

Critical Heat Flux Correlations, December 

2018, Revision 1, Docket: PROJ0769. 

22. TR–0316–22048–NP, Nuclear Steam 

Supply System Advanced Sensor Technical 

Report, May 2020, Revision 3, Docket: 52– 

048. 

23. TR–0515–13952–NP–A, Risk 

Significance Determination, October 2016, 

Revision 0, Docket: PROJ0769, NuScale 

Nonproprietary. 

24. TR–0516–49084–NP, Containment 

Response Analysis Methodology Technical 

Report, May 2020, Revision 3, Docket: 52– 

048. 

25. TR–0516–49416–NP–A, Non-Loss-ofCoolant Accident Analysis Methodology, July 

2020, Revision 3, Docket: PROJ0769. 


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26. TR–0516–49417–NP–A, Evaluation 

Methodology for Stability Analysis of the 

NuScale Power Module, March 2020, 

Revision 1, Docket: PROJ0769. 

27. TR–0516–49422–NP–A, Loss-ofCoolant Accident Evaluation Model, July 

2020, Revision 2, Docket: PROJ0769. 

28. TR–0616–48793–NP–A, Nuclear 

Analysis Codes and Methods Qualification, 

November 2018, Revision 1, Docket: 

PROJ0769. 

29. TR–0616–49121–NP, NuScale 

Instrument Setpoint Methodology Technical 

Report, May 2020, Revision 3, Docket: 52– 

048. 

30. TR–0716–50350–NP–A, Rod Ejection 

Accident Methodology, June 2020, Revision 

1, Docket: PROJ0769. 

31. TR–0716–50351–NP–A, NuScale 

Applicability of AREVA Method for the 

Evaluation of Fuel Assembly Structural 

Response to Externally Applied Forces, April 

2020, Revision 1, Docket: PROJ0769. 

32. TR–0716–50424–NP, Combustible Gas 

Control, March 2019, Revision 1, Docket: 

PROJ0769. 

33. TR–0716–50439–NP, NuScale 

Comprehensive Vibration Assessment 

Program Analysis Technical Report, July 

2019, Revision 2, Docket: 52–048. 

34. TR–0815–16497–NP–A, Safety 

Classification of Passive Nuclear Power Plant 

Electrical Systems, January 2018, Revision 1, 

Docket: PROJ0769. 

35. TR–0816–49833–NP, Fuel Storage Rack 

Analysis, November 2018, Revision 1, 

Docket: 52–048. 

36. TR–0816–50796–NP, Loss of Large 

Areas Due to Explosions and Fires 

Assessment, June 2019, Revision 1, Docket: 

52–048. 

37. TR–0816–50797, Mitigation Strategies 

for Loss of All AC Power Event, October 

2019, Revision 3, Docket: 52–048, NuScale 

Nonproprietary. 

38. TR–0816–51127–NP, NuFuel-HTP2TM

Fuel and Control Rod Assembly Designs, 

December 2019, Revision 3, Docket: 52–048. 

39. TR–0818–61384–NP, Pipe Rupture 

Hazards Analysis, July 2019, Revision 2, 

Docket No.: 52–048. 

40. TR–0915–17564–NP–A, Subchannel 

Analysis Methodology, February 2019, 

Revision 2, Docket: PROJ0769. 

41. TR–0915–17565–NP–A, Accident 

Source Term Methodology, February 2020, 

Revision 4, Docket: PROJ0769. 

42. TR–0916–51299–NP, Long-Term 

Cooling Methodology, May 2020, Revision 3, 

Docket: 52–048. 

43. TR–0916–51502–NP, NuScale Power 

Module Seismic Analysis, April 2019, 

Revision 2, Docket: 52–048. 

44. TR–0917–56119–NP, CNV Ultimate 

Pressure Integrity, June 2019, Revision 1, 

Docket No. 52–048. 

45. TR–0918–60894–NP, NuScale 

Comprehensive Vibration Assessment 

Program Measurement and Inspection Plan 

Technical Report, August 2019, Revision 1, 

Docket No.: 52–048. 

46. NP–TR–1010–859–NP–A, NuScale 

Topical Report: Quality Assurance Program 

Description for the NuScale Power Plant, 

May 2020, Revision 5, Docket: PROJ0769, 

NuScale Nonproprietary. 

47. TR–1015–18177–NP, Pressure and 

Temperature Limits Methodology, October 

2018, Revision 2, Docket: 52–048. 

48. TR–1015–18653–NP–A, Design of the 

Highly Integrated Protection System 

Platform, May 2017, Revision 2, Docket: 

PROJ0769. 

49. TR–1016–51669–NP, NuScale Power 

Module Short-Term Transient Analysis, July 

2019, Revision 1, Docket: 52–048. 

50. TR–1116–51962–NP, NuScale 

Containment Leakage Integrity Assurance, 

May 2019, Revision 1, Docket: 52–048. 

51. TR–1116–52065–NP, Effluent Release 

(GALE Replacement) Methodology and 

Results, November 2018, Revision 1, Docket: 

52–048. 

B.1. An applicant or licensee referencing 

this appendix, in accordance with Section IV 

of this appendix, shall incorporate by 

reference and comply with the requirements 

of this appendix except as otherwise 

provided in this appendix. 

2. Conceptual design information, as set 

forth in the design certification application 

Part 2, Tier 2, Section 1.2, and the discussion 

of ‘‘first principles’’ contained in design 

certification application Part 2, Tier 2, 

Section 14.3.2 are not incorporated by 

reference into this appendix. 

C. If there is a conflict between Tier 1 and 

Tier 2 of the DCD, then Tier 1 controls. 

D. If there is a conflict between the generic 

DCD and either the application for the design 

certification of NuScale or the final safety 

evaluation report related to certification of 

the NuScale standard design, then the 

generic DCD controls. 

E. Design activities for structures, systems, 

and components that are entirely outside the 

scope of this appendix may be performed 

using site characteristics, provided the design 

activities do not affect the DCD or conflict 

with the interface requirements. 

IV. Additional Requirements and 

Restrictions 

A. An applicant for a COL that wishes to 

reference this appendix shall, in addition to 

complying with the requirements of §§ 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 § 52.47(a) that is 

not within the scope of this appendix; 

g. Information demonstrating that 

necessary shielding to limit radiological dose 

consistent with the radiation zones specified 

in design certification application Part 2, Tier 

2, Chapter 12, Figure 12.3–1, ‘‘Reactor 

Building Radiation Zone Map,’’ is provided 

to account for penetrations in the radiation 

shield wall between the power module bay 

and the reactor building steam gallery area; 

h. Information demonstrating that the 

requirements of 10 CFR 50.34(f)(2)(xxviii) are 

met with respect to potential radiological 

releases under accident conditions from the 

systems used for post-accident hydrogen and 

oxygen monitoring described in design 

certification application Part 2, Tier 2, 

Section 6.2.5; information demonstrating that 

post-accident leakage from these systems 

does not result in the total main control room 

dose exceeding the dose criteria for the 

surrogate event with significant core damage, 

which may include use of design features 

compliant with 10 CFR 50.34(f)(2)(vii), as 

appropriate; and information demonstrating 

that post-accident leakage from these systems 

does not result in the total dose for the 

surrogate event with significant core damage 

exceeding the offsite dose criteria, as 

required by 10 CFR 52.47(a)(2)(iv); and 

i. Information demonstrating that the 

criteria of 10 CFR part 20 and the 

requirements of 10 CFR part 50, appendix A, 

General Design Criterion (GDC) 4 and GDC 31 

are met with respect to the structural and 

leakage integrity of the steam generator tubes 

that might be compromised by effects from 

density wave oscillations in the secondary 

fluid system, including the method of 

analysis to predict the thermal-hydraulic 

conditions of the steam generator secondary 

fluid system and resulting loads, stresses, 

and deformations from density wave 

oscillations and reverse flow. This 

information must be consistent with the 

other design information regarding steam 

generator integrity contained in design 

certification application Part 2, Tier 2, 

Sections 3.9.2 and 5.4.1. 

3. Include, in the plant-specific DCD, the 

sensitive, unclassified, non-safeguards 

information (including proprietary 

information and security-related information) 

and safeguards information referenced in the 

NuScale generic DCD. 

4. Include, as part of its application, a 

demonstration that an entity other than 

NuScale Power, LLC, is qualified to supply 

the NuScale generic DCD, unless NuScale 

Power, LLC, supplies the design for the 

applicant’s use. 

B. The Commission reserves the right to 

determine in what manner this appendix 

may be referenced by an applicant for a 

construction permit or operating license 

under 10 CFR part 50. 

V. Applicable Regulations 

A. Except as indicated in paragraph B of 

this section, the regulations that apply to 

NuScale are in 10 CFR parts 20, 50, 52, 73, 

and 100, codified as of [DATE 120 DAYS 

AFTER DATE OF PUBLICATION OF FINAL 

RULE IN THE Federal Register], that are 

applicable and technically relevant, as 

described in the final safety evaluation 

report. 


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B. The NuScale design is exempt from 

portions of the following regulations: 

1. Paragraph (f)(2)(vi) of 10 CFR 50.34 and 

10 CFR 50.46a—High point venting for the 

reactor coolant system and reactor pressure 

vessel head. 

2. Paragraph (f)(2)(viii) of 10 CFR 50.34— 

Post-accident sampling of the reactor coolant 

system and containment. 

3. Paragraph (f)(2)(xiii) of 10 CFR 50.34— 

Power supplies for pressurizer heaters. 

4. Paragraph (f)(2)(xiv)(E) of 10 CFR 

50.34—Automatic closing of containment 

isolation systems on a high radiation signal. 

5. Paragraph (f)(2)(xx) of 10 CFR 50.34— 

Power from vital buses and emergency power 

sources for pressurizer level indication. 

6. Paragraph (c)(2) of 10 CFR 50.44— 

Combustible gas control. 

7. Paragraph (a)(1)(i) of 10 CFR 50.46— 

Applicability limited to reactor designs that 

use zircaloy or ZIRLO fuel rod cladding 

material. 

8. Paragraph (m) of 10 CFR 50.54— 

Minimum Staffing. In lieu of these 

requirements, a licensee that references this 

appendix must comply with the following: 

a. A senior operator licensed pursuant to 

part 55 of this chapter shall be present at the 

facility or readily available on call at all 

times during its operation, and shall be 

present at the facility during initial startup 

and approach to power, recovery from an 

unplanned or unscheduled shutdown or 

significant reduction in power, and refueling, 

or as otherwise prescribed in the facility 

license. 

b. Licensees shall meet the following 

requirements: 

i. Each licensee shall meet the minimum 

licensed operator staffing requirements in the 

following table: 

TABLE 1—MINIMUM REQUIREMENTS PER SHIFT FOR ON-SITE STAFFING OF NUSCALE POWER PLANTS BY OPERATORS AND

SENIOR OPERATORS LICENSED UNDER 10 CFR PART 55 

Number of units operating 

(a nuclear power unit is considered to be operating when it is in MODE 1, 2, or 3 as defined by the 

unit’s technical specifications) 

Position 

One to twelve 

units 

One control 

room 

None ............................................................................................................................................................ Senior operator ..........

Operator ....................

1 

2 

One to twelve ............................................................................................................................................... Senior operator .........

Operator ....................

3 

3 

Source: Design Certification Application, Part 7, Section 6.1.3, ‘‘Requested Action.’’ 

ii. Each facility licensee shall have at its 

site a person holding a senior operator 

license for all fueled units at the site who is 

assigned responsibility for overall plant 

operation at all times there is fuel in any 

unit. At all times any module is fueled, 

regardless of Mode, there must be a licensed 

operator or senior operator in the control 

room. 

iii. When a nuclear power unit is in MODE 

1, 2, or 3, as defined by the unit’s technical 

specifications, each licensee shall have a 

person holding a senior operator license for 

the nuclear power unit in the control room 

at all times. In addition to this senior 

operator, a second person who is either a 

licensed operator or licensed senior operator 

shall be present at the controls at all times. 

A third person who is either a licensed 

operator or licensed senior operator shall be 

in the control room envelope at all times. 

iv. Each licensee shall have present, during 

alteration or movement of the core of a 

nuclear power unit (including fuel loading, 

fuel transfer, or movement of a module that 

contains fuel), a person holding a senior 

operator license or a senior operator license 

limited to fuel handling to directly supervise 

the activity and, during this time, the 

licensee shall not assign other duties to this 

person. 

9. Paragraph (c)(1) of 10 CFR 50.62— 

Diverse equipment to initiate a turbine trip 

under conditions indicative of an anticipated 

transient without scram. 

10. Appendix A of 10 CFR part 50— 

Electric Power Systems GDCs: 

a. GDC 17—Electric power systems for 

safety-related functions; 

b. GDC 18—Design to permit periodic 

inspection and testing of electric power 

systems; 

c. GDC 34—Electric power systems for 

residual heat removal; 

d. GDC 35—Electric power systems for 

emergency core cooling; 

e. GDC 38—Electric power systems for 

containment heat removal; 

f. GDC 41—Electric power systems for 

containment atmosphere cleanup; and 

g. GDC 44—Electric power systems for 

cooling. 

11. Appendix A to 10 CFR part 50, GDC 

19—Equipment outside the control room 

with capability for cold shutdown of the 

reactor. 

12. Appendix A to 10 CFR part 50, GDC 

27—Demonstration of long-term shutdown 

under post-accident conditions with an 

assumed worst rod stuck out. 

13. Appendix A to 10 CFR part 50, GDC 

33—Reactor coolant makeup for protection 

against small breaks in the reactor coolant 

pressure boundary. 

14. Appendix A to 10 CFR part 50, GDC 

40—Periodic pressure and functional testing 

of containment heat removal system. 

15. Appendix A to 10 CFR part 50, GDC 

52—Design to allow periodic containment 

leakage rate testing. 

16. Appendix A of 10 CFR part 50, GDCs 

55, 56, and 57—Containment Isolation: 

a. GDC 55—Isolation valves for certain 

reactor coolant pressure boundary lines 

penetrating containment; 

b. GDC 56—Isolation valves for certain 

primary containment lines; and 

c. GDC 57—Isolation valves for certain 

closed systems lines. 

17. Appendix K to 10 CFR part 50— 

Emergency Core Cooling System Evaluation 

Models: 

a. Section I.A.4—Heat generation rates 

from radioactive decay of fission products; 

b. Section I.A.5—Rate of energy release, 

hydrogen generation, and cladding oxidation 

from the metal/water reaction; 

c. Section I.B—Predicting cladding 

swelling and rupture; 

d. Section I.C.1.b—Calculation of the 

discharge rate for all times after the 

discharging fluid has been calculated to be 

two-phase; 

e. Section I.C.5.a—Post-critical heat flux 

correlations of heat transfer from the fuel 

cladding to the surrounding fluid; and 

f. Section I.C.7.a—Calculation of cross-flow 

between the hot and average channel regions 

of the core during blowdown. 

VI. Issue Resolution 

A. The Commission has determined that 

the structures, systems, and components and 

design features of NuScale comply with the 

provisions of the Atomic Energy Act of 1954, 

as amended, and the applicable regulations 

identified in Section V of this appendix; and 

therefore, provide adequate protection to the 

health and safety of the public. A conclusion 

that a matter is resolved includes the finding 

that additional or alternative structures, 

systems, and components, design features, 

design criteria, testing, analyses, acceptance 

criteria, or justifications are not necessary for 

NuScale. 

B. The Commission considers the 

following matters resolved within the 

meaning of § 52.63(a)(5) in subsequent 

proceedings for issuance of a COL, 

amendment of a COL, or renewal of a COL, 

proceedings held under § 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; 


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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 thermalhydraulic conditions of the steam generator 

secondary fluid system and resulting loads, 

stresses, and deformations from density wave 

oscillations and reverse flow, consistent with 

the other design information regarding steam 

generator integrity described in DCA Part 2, 

Tier 2, Sections 3.9.1, 3.9.2, 5.4.1, and 15.6.3, 

and in accordance with 10 CFR part 50, GDC 

4, 10, and 31; 

2. All nuclear safety and safeguards issues 

associated with the referenced information in 

the non-public documents in Tables 1.6–1 

and 1.6–2 of Tier 2 of the DCD, which 

contain sensitive unclassified non-safeguards 

information (including proprietary 

information and security-related information) 

and safeguards information and which, in 

context, are intended as requirements in the 

generic DCD for the NuScale design; 

3. All generic changes to the DCD under 

and in compliance with the change processes 

in paragraphs VIII.A.1 and VIII.B.1 of this 

appendix; 

4. All exemptions from the DCD under and 

in compliance with the change processes in 

paragraphs VIII.A.4 and VIII.B.4 of this 

appendix, but only for that plant; 

5. All departures from the DCD that are 

approved by license amendment, but only for 

that plant; 

6. Except as provided in paragraph 

VIII.B.5.g of this appendix, all departures 

from Tier 2 under and in compliance with 

the change processes in paragraph VIII.B.5 of 

this appendix that do not require prior NRC 

approval, but only for that plant; and 

7. All environmental issues concerning 

severe accident mitigation design alternatives 

associated with the information in the NRC’s 

environmental assessment for NuScale 

(ADAMS Accession No. ML19303C179) and 

DCD Part 3, ‘‘Applicant’s Environmental 

Report—Standard Design Certification,’’ 

Revision 5, dated July 2020 (ADAMS 

Accession No. ML20224A512), for plants 

referencing this appendix whose site 

characteristics fall within those site 

parameters specified in the NuScale 

environmental report. 

C. The Commission does not consider 

operational requirements for an applicant or 

licensee who references this appendix to be 

matters resolved within the meaning of 

§ 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 § 52.85, the hearing provided 

under § 52.103, or in any other proceeding 

relating to this appendix, in which interested 

persons have a right to request an 

adjudicatory hearing. 

VII. Duration of This Appendix 

This appendix may be referenced for a 

period of 15 years from October 29, 2021, 

except as provided for in §§ 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 

§ 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 § 52.63(a)(4). 

4. Exemptions from Tier 1 information are 

governed by the requirements in 

§§ 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 

§ 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 plantspecific order, while this appendix is in 

effect under § 52.55 or § 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 plantspecific 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 

§ 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 plantspecific DCD; 


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(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 exvessel severe accident previously reviewed. 

d. A proposed departure from Tier 2 

information required by § 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 

§ 52.103(a), who believes that an applicant or 

licensee who references this appendix has 

not complied with paragraph VIII.B.5 of this 

appendix when departing from Tier 2 

information, may petition to admit into the 

proceeding such a contention. In addition to 

complying with the general requirements of 

10 CFR 2.309, the petition must demonstrate 

that the departure does not comply with 

paragraph VIII.B.5 of this appendix. Further, 

the petition must demonstrate that the 

change stands on an asserted noncompliance 

with an ITAAC acceptance criterion in the 

case of a § 52.103 preoperational hearing, or 

that the change stands directly on the 

amendment request in the case of a hearing 

on a license amendment. Any other party 

may file a response. If, on the basis of the 

petition and any response, the presiding 

officer determines that a sufficient showing 

has been made, the presiding officer shall 

certify the matter directly to the Commission 

for determination of the admissibility of the 

contention. The Commission may admit such 

a contention if it determines the petition 

raises a genuine issue of material fact 

regarding compliance with paragraph VIII.B.5 

of this appendix. 

C. Operational Requirements 

1. Changes to NuScale design certification 

generic TS and other operational 

requirements that were completely reviewed 

and approved in the design certification rule 

and do not require a change to a design 

feature in the generic DCD are governed by 

the requirements in 10 CFR 50.109. Changes 

that require a change to a design feature in 

the generic DCD are governed by the 

requirements in paragraphs A or B of this 

section. 

2. Changes to NuScale design certification 

generic TS and other operational 

requirements are applicable to all applicants 

who reference this appendix, except those for 

which the change has been rendered 

technically irrelevant by action taken under 

paragraphs C.3 or C.4 of this section. 

3. The Commission may require plantspecific 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 § 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 § 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 § 2.309 of this 

chapter and must either demonstrate why 

special circumstances as defined in § 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 plantspecific 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 plantspecific DCD to accurately reflect both 

generic changes to the generic DCD and 

plant-specific departures made under Section 

VIII of this appendix throughout the period 

of application and for the term of the license 

(including any periods of renewal). 

3. An applicant or licensee who references 

this appendix shall prepare and maintain 

written evaluations which provide the bases 

for the determinations required by Section 

VIII of this appendix. These evaluations must 

be retained throughout the period of 

application and for the term of the license 

(including any periods of renewal). 

4.a. The applicant for NuScale shall 

maintain a copy of the aircraft impact 

assessment performed to comply with the 

requirements of 10 CFR 50.150(a) for the term 

of the certification (including any period of 

renewal). 

b. An applicant or licensee who references 

this appendix shall maintain a copy of the 

aircraft impact assessment performed to 

comply with the requirements of 10 CFR 

50.150(a) throughout the pendency of the 

application and for the term of the license 

(including any periods of renewal). 

B. Reporting 

1. An applicant or licensee who references 

this appendix shall submit a report to the 

NRC containing a brief description of any 

plant-specific departures from the DCD, 

including a summary of the evaluation of 

each departure. This report must be filed in 

accordance with the filing requirements 

applicable to reports in § 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. 


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Federal Register / Vol. 86, No. 124 / Thursday, July 1, 2021 / Proposed Rules 35023 

b. During the interval from the date of 

application for a license to the date the 

Commission makes its finding required by 

§ 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 § 52.103(g), the reports and 

updates to the plant-specific DCD must be 

submitted, along with updates to the sitespecific portion of the final safety analysis 

report for the facility, at the intervals 

required by 10 CFR 50.59(d)(2) and 

50.71(e)(4), respectively, or at shorter 

intervals as specified in the license. 

Dated: June 25, 2021. 

For the Nuclear Regulatory Commission. 

Annette Vietti-Cook, 

Secretary of the Commission. 

[FR Doc. 2021–13940 Filed 6–30–21; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Part 52 

[NRC–2017–0090] 

RIN 3150–AK04 

Advanced Boiling Water Reactor 

(ABWR) Design Certification Renewal 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Proposed rule and 

environmental assessment. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC) is proposing to 

amend its regulations to renew the U.S. 

Advanced Boiling Water Reactor 

standard design certification. 

Applicants or licensees intending to 

construct and operate a U.S. Advanced 

Boiling Water Reactor standard design 

may do so by referencing this design 

certification rule. The applicant for the 

renewal of the U.S. Advanced Boiling 

Water Reactor standard design 

certification is General Electric-Hitachi 

Nuclear Energy Americas, LLC. The 

NRC invites public comment on this 

proposed rule and environmental 

assessment. 

DATES: Submit comments by August 2, 

2021. Comments received after this date 

will be considered if it is practical to do 

so, but the NRC is able to ensure 

consideration only for comments 

received on or before this date. 

ADDRESSES: You may submit comments 

by any of the following methods (unless 

this document describes a different 

method for submitting comments on a 

specific subject): 

• Federal Rulemaking Website: Go to 

https://www.regulations.gov and search 

for Docket ID NRC–2017–0090. Address 

questions about NRC dockets to Dawn 

Forder; telephone: 301–415–3407; 

email: Dawn.Forder@nrc.gov. For 

technical questions contact the 

individuals listed in the FOR FURTHER

INFORMATION CONTACT section of this 

document. 

• Email comments to: 

Rulemaking.Comments@nrc.gov. If you 

do not receive an automatic email reply 

confirming receipt, then contact us at 

301–415–1677. 

For additional direction on obtaining 

information and submitting comments, 

see ‘‘Obtaining Information and 

Submitting Comments’’ in the 

SUPPLEMENTARY INFORMATION section of 

this document. 

FOR FURTHER INFORMATION CONTACT: 

Dennis Andrukat, Office of Nuclear 

Material Safety and Safeguards, 

telephone: 301–415–3561, email: 

Dennis.Andrukat@nrc.gov, or James 

Shea, Office of Nuclear Reactor 

Regulation, telephone: 301–415–1388, 

email: James.Shea@nrc.gov. Both are 

staff of the U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001. 

SUPPLEMENTARY INFORMATION: 

Table of Contents 

I. Obtaining Information and Submitting 

Comments 

II. Rulemaking Procedure 

III. Background 

IV. Voluntary Consensus Standards 

V. Plain Writing 

VI. Environmental Assessment and Final 

Finding of No Significant Impact 

VII. Paperwork Reduction Act Statement 

VIII. Availability of Documents 

I. Obtaining Information and 

Submitting Comments 

A. Obtaining Information 

Please refer to Docket ID NRC–2017– 

0090 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–2017–0090. 

• NRC’s Agencywide Documents 

Access and Management System 

(ADAMS): You may obtain publiclyavailable documents online in the 

ADAMS Public Documents collection at 

https://www.nrc.gov/reading-rm/ 

adams.html. To begin the search, select 

‘‘Begin Web-based ADAMS Search.’’ For 

problems with ADAMS, please contact 

the NRC’s Public Document Room (PDR) 

reference staff at 1–800–397–4209, at 

301–415–4737, or by email to 

PDR.Resource@nrc.gov. For the 

convenience of the reader, instructions 

about obtaining materials referenced in 

this document are provided in the 

Availability of Documents section. 

• Attention: The Public Document 

Room (PDR), where you may examine 

and order copies of public documents is 

currently closed. You may submit your 

request to the PDR via email at 

PDR.Resource@nrc.gov or call 1–800– 

397–4209 between 8:00 a.m. and 4:00 

p.m. (EST), Monday through Friday, 

except Federal holidays. 

• Attention: The Technical Library, 

which is located at Two White Flint 

North, 11545 Rockville Pike, Rockville, 

Maryland 20852, is open by 

appointment only. Interested parties 

may make appointments to examine 

documents by contacting the NRC 

Technical Library by email at 

Library.Resource@nrc.gov between 8:00 

a.m. and 4:00 p.m. (EST), Monday 

through Friday, except Federal holidays. 

B. Submitting Comments 

The NRC encourages electronic 

comment submission through the 

Federal Rulemaking Website (https://

www.regulations.gov). Please include 

Docket ID NRC–2017–0090 in your 

comment submission. 

The NRC cautions you not to include 

identifying or contact information that 

you do not want to be publicly 

disclosed in your comment submission. 

The NRC will post all comment 

submissions at https://

www.regulations.gov as well as enter the 

comment submissions into ADAMS. 

The NRC does not routinely edit 

comment submissions to remove 

identifying or contact information. 

If you are requesting or aggregating 

comments from other persons for 

submission to the NRC, then you should 

inform those persons not to include 

identifying or contact information that 

they do not want to be publicly 

disclosed in their comment submission. 

Your request should state that the NRC 

does not routinely edit comment 

submissions to remove such information 

before making the comment 

submissions available to the public or 

entering the comment into ADAMS. 

II. Rulemaking Procedure 

Because the NRC anticipates that this 

action will be non-controversial, the 

NRC is publishing this proposed rule 

concurrently with a direct final rule in 

the Rules and Regulations section of this 

issue of the Federal Register. The direct 

final rule will become effective on 

September 29, 2021. However, if the 

NRC receives significant adverse 


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