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Proposed Rule2026-08550

Licensing Requirements for Microreactors and Other Reactors With Comparable Risk Profiles

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Published

May 1, 2026

Issuing agencies

Nuclear Regulatory Commission

Abstract

The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its regulations to establish a risk-informed and performance- based regulatory framework for rapid licensing of new microreactors and other reactors with comparable risk profiles and for high-volume deployment of these reactors. The proposed rule would provide a flexible set of licensing pathways, reduce regulatory burden, and ensure that safety and security requirements remain commensurate with the potential hazards posed by these facilities.

Full Text

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[Federal Register Volume 91, Number 84 (Friday, May 1, 2026)]
[Proposed Rules]
[Pages 23628-23766]
From the Federal Register Online via the Government Publishing Office [<a href="http://www.gpo.gov">www.gpo.gov</a>]
[FR Doc No: 2026-08550]

[[Page 23627]]

Vol. 91

Friday,

No. 84

May 1, 2026

Part III

Nuclear Regulatory Commission

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10 CFR Parts 1, 2, 10, et. al.

Licensing Requirements for Microreactors and Other Reactors With
Comparable Risk Profiles; Proposed Rule

Federal Register / Vol. 91, No. 84 / Friday, May 1, 2026 / Proposed
Rules

[[Page 23628]]

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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 57,
70, 72, 73, 74, 75, 95, 140, 150

[NRC-2025-0379]
RIN 3150-AL36

Licensing Requirements for Microreactors and Other Reactors With
Comparable Risk Profiles

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule; guidance; and request for comment.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to
amend its regulations to establish a risk-informed and performance-
based regulatory framework for rapid licensing of new microreactors and
other reactors with comparable risk profiles and for high-volume
deployment of these reactors. The proposed rule would provide a
flexible set of licensing pathways, reduce regulatory burden, and
ensure that safety and security requirements remain commensurate with
the potential hazards posed by these facilities.

DATES: Comments must be submitted electronically using <a href="https://www.regulations.gov">https://www.regulations.gov</a> by 11:59 p.m. eastern time on June 15, 2026.

ADDRESSES: Submit your comments, identified by Docket ID NRC-2025-0379,
at <a href="https://www.regulations.gov">https://www.regulations.gov</a>. If your material cannot be submitted
using <a href="https://www.regulations.gov">https://www.regulations.gov</a>, call or email the individuals listed
in the FOR FURTHER INFORMATION CONTACT section of this document for
alternate instructions.
Do not include any personally identifiable information (such as
name, address, or other contact information) or confidential business
information that you do not want publicly disclosed. All comments are
public records; they are publicly displayed exactly as received, and
will not be deleted, modified, or redacted. Comments may be submitted
anonymously.
Follow the search instructions on <a href="https://www.regulations.gov">https://www.regulations.gov</a> to
view public comments.
You can read a plain language description of this proposed rule at
<a href="https://www.regulations.gov/docket/NRC-2025-0379">https://www.regulations.gov/docket/NRC-2025-0379</a>. 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: George Tartal, Office of Nuclear
Material Safety and Safeguards, telephone: 301-415-0016, email:
<a href="/cdn-cgi/l/email-protection#2562404a5742400b714457514449654b57460b424a53">[email&#160;protected]</a>; Elijah Dickson, Office of Nuclear Reactor
Regulation, telephone: 301-415-7647, email: <a href="/cdn-cgi/l/email-protection#2d684144474c450369444e465e42436d435f4e034a425b">[email&#160;protected]</a>;
Michael Balazik, Office of Nuclear Reactor Regulation, telephone: 301-
415-2856, email: <a href="/cdn-cgi/l/email-protection#0944606a61686c65274b68656873606249677b6a276e667f">[email&#160;protected]</a>; and William Kennedy,
telephone: 301-415-2313, email: <a href="/cdn-cgi/l/email-protection#3f68565353565e5211745a51515a5b467f514d5c11585049">[email&#160;protected]</a>. All are staff
of the U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

SUPPLEMENTARY INFORMATION:

Executive Summary

A. Need for the Regulatory Action

The purpose of this rulemaking is to safely expedite the licensing
process for microreactors and other reactors with comparable risk
profiles. This effort is consistent with, and implements direction in,
the Accelerating Deployment of Versatile, Advanced Nuclear for Clean
Energy Act of 2024 (Pub. L. 118-67, 138 Stat. 1448) (ADVANCE Act), and
Executive Order (E.O.) 14300, ``Ordering the Reform of the Nuclear
Regulatory Commission'' (90 FR 22587; May 29, 2025).
Section 208 of the ADVANCE Act requires the NRC to develop ``risk-
informed and performance-based strategies and guidance to license and
regulate microreactors.'' The ADVANCE Act mandates that these
strategies be incorporated into the existing regulatory framework, the
technology-inclusive regulatory framework to be established through the
rulemaking required by section 103(a)(4) of the Nuclear Energy
Innovation and Modernization Act (Pub. L. 115-439, 132 Stat. 5572)
(NEIMA), or a pending or new rulemaking by July 2027.
On January 20, 2025, the President declared a National Energy
Emergency in E.O. 14156, ``Declaring a National Energy Emergency'' (90
FR 8433; January 29, 2025), and stressed the need for a reliable,
diversified, and affordable supply of energy. The President also issued
E.O. 14154 (90 FR 8353; January 29, 2025), titled, ``Unleashing
American Energy,'' with an objective of unleashing ``America's
affordable and reliable energy and natural resources.''
On May 23, 2025, the President issued E.O. 14300. Section 5(e) of
that E.O. directs the NRC to revise its regulations to ``[e]stablish a
process for high-volume licensing of microreactors and modular
reactors, including by allowing for standardized applications and
approvals and by considering to what extent such reactors or components
thereof should be regulated through general licenses.'' That E.O. set
February 23, 2026, as the deadline for issuing this proposed rule, and
the final rule must be issued by November 23, 2026.
In developing this proposed rule, the NRC considered whether to
establish the rule's scope within the amended non-power production or
utilization facility (NPUF) licensing framework set out in the NRC's
final rule, ``Non-Power Production or Utilization Facility License
Renewal,'' issued on December 30, 2024 (89 FR 106234). That NPUF
rulemaking was primarily intended to revise and streamline the license
renewal process for facilities such as research and test reactors and
medical isotope production facilities and was not designed to serve as
a comprehensive licensing pathway for the high-volume deployment of
microreactors. However, many of the design features and siting
characteristics of NPUFs are expected to closely align with those
reactors within the scope of this rulemaking. NPUFs are commonly
located at national laboratories, private ventures, and universities,
situated in both sparsely and densely populated areas. They operate
over a broad range of thermal powers--up to tens of megawatts--with
large thermal capacities and fuel designed with inherent safety
features that enhance their stability and safety.
The NRC considered amending part 50, ``Domestic Licensing of
Production and Utilization Facilities,'' or part 52, ``Licenses,
Certifications, and Approvals For Nuclear Power Plants,'' of title 10
of the Code of Federal Regulations (10 CFR), to provide for high-volume
licensing of microreactors and other reactors with comparable risk
profiles. The NRC didn't pursue amending part 52 or implementing a
combined license approach in this proposed rule because the
requirements for inspections, tests, analyses, and acceptance criteria
(ITAAC) were designed for light water reactors (LWRs) (required by the
Atomic Energy Act of 1954, as amended (AEA)) and the associated hearing
on ITAAC closure could extend the licensing timeline. The NRC didn't
pursue amending part 50 because the regulations in part 50 for
commercial reactors were designed for large LWRs.
The NRC also considered developing this proposed rule's scope
within the framework of 10 CFR part 53, ``Risk-Informed, Technology-
Inclusive Regulatory Framework for Commercial Nuclear Plants.''
Although part 53 provides a pathway to support licensing of
microreactors, part 53 is designed to also cover large, complex
reactors. The

[[Page 23629]]

NRC decided to create a new part in 10 CFR chapter I that would be
focused on rapid and high-volume licensing of microreactors and other
reactors with comparable risk profiles. Therefore, the NRC developed a
separate rulemaking that combines elements of the Commission's NPUF
licensing approach in 10 CFR part 50 with elements from 10 CFR parts 52
and 53 to create proposed part 57, ``Licensing Requirements for
Microreactors and Other Reactors with Comparable Risk Profiles.'' This
proposed rule's framework would support rapid licensing of first-of-a-
kind microreactors and other reactors with comparable risk profiles and
high-volume deployment of these reactors through multiple licensing
pathways, including the option for a general license to construct parts
of these facilities.
Collectively, the NRC's regulatory frameworks offer optionality and
enable applicants to select licensing pathways that align with
applicant-specific circumstances and deployment strategies.

B. Major Provisions

The primary provisions of this proposed rule would establish a
risk-informed and performance-based regulatory framework for rapid and
high-volume licensing of microreactors and reactors with comparable
risk profiles. The proposed rule would provide flexible licensing
pathways with streamlined requirements, as compared to the analogous
requirements in part 50 and part 52, that would ensure safety and
security requirements remain commensurate with the potential hazards
posed by these facilities. Licensing and approval pathways would
include a construction permit (CP) and an operating license (OL), a
manufacturing license, a standard design approval, and provisions for
affording regulatory finality to nuclear plant designs and essentially
complete standardized operational programs. Applicants could combine in
a single application requests for these licenses and approvals with
requests for other licenses, approvals, and certifications for special
nuclear material, byproduct material, transportation, and irradiated
fuel storage to enable a broad spectrum of deployment models.
The proposed rule is intended to expedite licensing reviews based
on the statutory requirements of the AEA. E.O. 14300 directs the NRC to
reach a final decision on an application to construct and operate a new
reactor of any type within 18 months. This proposed licensing process
should enable the NRC to issue an OL within 6-12 months after accepting
an application, assuming that several factors beyond the NRC's control
are met (e.g., the application contains adequate information to allow
the NRC to immediately docket the application and does not require the
NRC to issue requests for additional information, the licensee
completes timely construction, and any hearing contentions are
expeditiously resolved). For a joint application for a CP and
associated OL(s), the applicant would be required to submit final
design information and complete operational programs at the time of
application. The NRC would conduct a single, comprehensive safety
review and potentially hold one adjudicatory hearing on the joint
application. The Advisory Committee on Reactor Safeguards would review
each joint application, focusing on aspects of the design that are
unique, novel, and noteworthy.
This proposed licensing framework would contain performance-based
and risk-informed entry criteria consistent with design attributes that
are necessary and essential for rapid, high-volume licensing of
microreactors and other reactors with comparable risk profiles.
Flexibilities in the proposed rule would include allowing a graded site
characterization approach using existing site characterization data
from Federal, State, or other organizations, provided that the data
meets applicable NRC quality standards. Also, applicants would be able
to define certain regulatory terms (e.g., ``basic component'' and
``safety-related'') and to limit the definition of ``construction'' to
safety-related structures, systems, and components (SSCs), as defined
in the proposed rule, or SSCs that would be relied upon to implement
the proposed security requirements.
The proposed rule would provide applicants with other
flexibilities. Applicants could propose and justify an appropriate use
of codes and standards as well as quality assurance programs tailored
to the safety significance of the facility's SSCs. For environmental
reviews, the proposed rule would permit the use of categorical
exclusions under the National Environmental Policy Act, provided that
specific conditions are met. The proposed rule would provide a general
license for certain construction activities before issuance of a CP for
an ``nth-of-a-kind'' facility (i.e., a nuclear reactor or nuclear plant
of a design that the NRC has already approved in a licensing
proceeding) if certain conditions are met. The proposed rule would also
provide alternative fitness-for-duty requirements for these licenses,
as well as require the development of a cybersecurity program using a
consequence-based approach.

C. Costs and Benefits

The NRC prepared a draft regulatory analysis to determine the
expected quantitative costs and benefits of this proposed rule and
associated guidance as well as qualitative factors to be considered in
the NRC's rulemaking decision. The conclusion from the analysis is that
this proposed rule and associated guidance would result in net averted
costs to the industry and the NRC of approximately $3.76 billion using
a 7-percent discount rate and $11.84 billion using a 3-percent discount
rate. As the number of applicants increases, so do the estimated
averted costs.
The draft regulatory analysis also considers qualitative factors,
such as greater regulatory stability, predictability, and clarity to
the licensing process. Another qualitative factor is promoting a
performance-based regulatory framework that specifies requirements to
be met and provides flexibility to an applicant or licensee regarding
the information or approach needed to satisfy those requirements.
For more information, please see the draft regulatory analysis
(available in the NRC's Agencywide Documents Access and Management
System (ADAMS) Accession No. ML26111A076).

Table of Contents

I. Obtaining Information and Submitting Comments
A. Obtaining Information
B. Submitting Comments
II. Executive Order 14300: Ordering the Reform of the Nuclear
Regulatory Commission
III. Background
A. Characteristics of Microreactors and Other Reactors With
Comparable Risk Profiles
B. Public Interest in Microreactors and Other Reactors With
Comparable Risk Profiles
IV. Discussion
A. Need for an Alternative Regulatory Framework
B. Description of Proposed Licensing Framework
C. Utilization Facilities and General Licenses
V. Part 57 Framework
A. Discussion of Provisions in Proposed Part 57
B. Subpart A--General Provisions
C. Subpart B--Eligibility
D. Subpart C--Construction Permits and Operating Licenses
E. Subpart D--Manufacturing Licenses
F. Subpart E--Standard Design Approvals

[[Page 23630]]

G. Subpart F--Reporting of Defects and Noncompliance
H. Subpart G--Irradiated Fuel Storage, Decommissioning, and
License Termination Requirements
I. Subpart H--Maintaining and Revising Licensing Basis
Information
J. Subpart I--Transportation Package Design Certification
K. Subpart J--Physical Security Requirements
L. Subpart K--Categorical Exclusion
M. Subpart L--Inspections
N. Subpart M--Material Control and Accounting
O. Subpart N--[Reserved]
P. Subpart O--Enforcement
Q. Subpart P--Operator Licensing and Human Factors
R. Subpart Q--Reporting and Other Administrative Requirements
VI. Changes to Other Parts of 10 CFR Chapter I
A. Conforming Changes to 10 CFR Parts 1, 2, 10, 11, 19, 20, 21,
25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, and 150
B. 10 CFR Part 26
C. 10 CFR Part 73
D. 10 CFR Part 140
VII. Specific Requests for Comments
VIII. Regulatory Flexibility Certification
IX. Regulatory Analysis
X. Backfitting and Issue Finality
XI. Cumulative Effects of Regulation
XII. Plain Writing
XIII. Environmental Assessment and Proposed Finding of No
Significant Environmental Impact
A. Introduction
B. Conforming Changes
C. Environmental Impacts of the Proposed Action
D. Environmental Impacts of the Alternative to the Proposed
Agency Action
E. Agencies and Persons Consulted
F. Proposed Finding of No Significant Environmental Impacts
G. Stakeholder Interactions
H. Environmental Assessment References
XIV. Paperwork Reduction Act
XV. Executive Orders
A. Executive Order 12866: Regulatory Planning and Review (as
Amended by Executive Order 14215, Ensuring Accountability for All
Agencies)
B. Executive Order 14154: Unleashing American Energy
C. Executive Order 14192: Unleashing Prosperity Through
Deregulation
D. Executive Order 14270: Zero-Based Regulatory Budgeting To
Unleash American Energy
E. Executive Order 14294: Fighting Overcriminalization in
Federal Regulations
XVI. Voluntary Consensus Standards
XVII. Availability of Guidance
XVIII. Public Meeting
XIX. Availability of Documents

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2025-0379 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:
<bullet> Federal Rulemaking Website: Go to <a href="https://www.regulations.gov">https://www.regulations.gov</a> and search for Docket ID NRC-2025-0379.
<bullet> NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly available documents online in the
ADAMS Public Documents collection at <a href="https://www.nrc.gov/reading-rm/adams.html">https://www.nrc.gov/reading-rm/adams.html</a>. 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 <a href="/cdn-cgi/l/email-protection#1d4d594f334f786e72686f7e785d736f7e337a726b">[email&#160;protected]</a>. For the convenience of the reader,
instructions about obtaining materials referenced in this document are
provided in the ``Availability of Documents'' section.
<bullet> NRC's PDR: The PDR, where you may examine and order copies
of publicly available documents, is open by appointment. To make an
appointment to visit the PDR, please send an email to
<a href="/cdn-cgi/l/email-protection#eebeaabcc0bc8b9d819b9c8d8bae809c8dc0898198">[email&#160;protected]</a> or call 1-800-397-4209 or 301-415-4737, between 8
a.m. and 4 p.m. eastern time, Monday through Friday, except Federal
holidays.
<bullet> Public Meeting: The NRC may conduct a public meeting to
describe the proposed amendments and answer questions from the public
on the proposed rule. If the NRC determines it will hold a public
meeting, NRC will publish a notice of the location, time, and agenda of
the meeting on the NRC's public meeting website within 10 calendar days
of the meeting. Stakeholders should monitor the NRC's public meeting
website for information about the public meeting at: <a href="https://www.nrc.gov/public-involve/public-meetings/index.cfm">https://www.nrc.gov/public-involve/public-meetings/index.cfm</a>.

B. Submitting Comments

Comments must be submitted electronically using <a href="https://www.regulations.gov">https://www.regulations.gov</a> by 11:59 p.m. eastern time on June 15, 2026. Please
include Docket ID NRC-2025-0379 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
<a href="https://www.regulations.gov">https://www.regulations.gov</a> 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. Executive Order 14300: Ordering the Reform of the Nuclear
Regulatory Commission

On May 23, 2025, President Donald J. Trump signed Executive Order
(E.O.) 14300, ``Ordering the Reform of the Nuclear Regulatory
Commission.'' Section 5, ``Reforming and Modernizing the NRC's
Regulations,'' requires the NRC to undertake a review and wholesale
revision of its regulations and guidance documents as guided by the
policies set forth in section 2 of the E.O. This rulemaking addresses
section 5(e), which requires the NRC to ``[e]stablish a process for
high-volume licensing of microreactors and modular reactors, including
by allowing for standardized applications and approvals and by
considering to what extent such reactors or components thereof should
be regulated through general licenses.''

III. Background

A. Characteristics of Microreactors and Other Reactors With Comparable
Risk Profiles

The microreactors and other reactors with comparable risk profiles
that would be licensed under this proposed rule would be commercial
nuclear reactors under section 103, ``Commercial Licenses,'' of the
Atomic Energy Act of 1954, as amended (AEA). Due to their expected
small sizes, low power levels, potential mobility, and simplicity of
operation compared to the current fleet of operating power reactors,
microreactors and other reactors with comparable risk profiles may be
useful, for example, for remote communities, non-electric industrial
processes, military bases, maritime applications, disaster relief, and
other applications where a grid connection is unreliable or
nonexistent.
Microreactors and other reactor concepts with comparable risk
profiles encompass a wide variety of reactor designs, including fuel
forms, coolant types, and power levels. These concepts often
incorporate inherent and passive safety design features that
distinguish them from the large light water reactors

[[Page 23631]]

in the current operating fleet. Fuel forms vary widely, from
traditional light water reactor fuel assemblies to advanced fuels such
as tri-structural isotropic (TRISO) particles, metallic fuels, and
liquid fuels. Coolants include water, liquid metals (e.g., sodium,
lead), inert gases (e.g., helium), and various molten salts. Power
outputs range from only a few kilowatts to several tens of megawatts,
and designs may operate in either a fast or thermal neutron spectrum.
These diverse technical approaches reflect the industry's pursuit of
reactor systems optimized for specific missions, operational
environments, and market applications.
Based on input from stakeholders (see section III.B, ``Public
Interest in Microreactors and Other Reactors with Comparable Risk
Profiles,'' of this document), the NRC anticipates that microreactors
and other reactors with comparable risk profiles would rely heavily on
standardization of design features and mass production to simplify
licensing and deployment. Some reactors may be ``self-contained'' in
that they would incorporate the reactor, shielding, and balance of
plant in one or several transportable containers and require minimal
site preparation or construction activities at the deployment site.
Other designs may consist of a nuclear reactor that would be fabricated
in a manufacturing facility and then incorporated into or connected to
the permanent structures and systems of a nuclear plant constructed at
the deployment site, such as a reactor building and power conversion
equipment.
The NRC understands that deployment models for microreactors and
other reactors with comparable risk profiles would include various
activities involving NRC licensing, certification, or approval. These
activities may include designing reactors, manufacturing at a
manufacturing facility, loading fuel at a manufacturing facility,
operating the reactors for testing at a manufacturing facility,
transporting fueled reactors to deployment sites (loaded with
unirradiated or irradiated fuel), operating the reactors for the
production of electrical or heat energy at the deployment sites,
replacing reactors at the deployment sites, transporting reactors away
from the deployment sites at the end of their useful lives,
decommissioning or refurbishing and refueling reactors at locations
away from the deployment sites, and re-deploying refurbished reactors
to deployment sites. Some microreactors and other reactors with
comparable risk profiles may also use more ``traditional'' approaches,
including constructing the reactor in its entirety, loading fuel, or
performing operational testing at the deployment site. This proposed
rule would provide processes and requirements that would enable all
these potential deployment models.

B. Public Interest in Microreactors and Other Reactors With Comparable
Risk Profiles

The NRC recognizes the public interest in the development and
deployment of microreactors and other reactors with comparable risk
profiles. For several years, the NRC has conducted advanced reactor
stakeholder meetings to facilitate open communication between the
agency, industry, and the public regarding regulatory policy, licensing
pathways, and technical issues related to advanced reactors. These
meetings covered a wide range of topics, including safety and security
considerations, fuel qualification and transportation, siting and
environmental review, emergency preparedness, quality assurance
approaches, risk-informed and performance-based regulatory methods, and
lessons learned from the licensing of non-power production or
utilization facilities (NPUFs). Stakeholders have also discussed and
presented strategies for streamlining licensing processes to
accommodate the anticipated high licensing volumes associated with
modular and transportable reactor concepts.
In addition to these public meetings, the NRC has received letters
and formal reports from a broad spectrum of interested parties,
including non-governmental organizations, policy organizations
representing both the nuclear industry and public interest groups,
national laboratories, and Federal, State, and local governmental
entities. These submissions have provided perspectives on technical
design features, operational considerations, safety analysis
methodologies, environmental impacts, workforce development, and policy
objectives for advanced reactor deployment. Many communications have
highlighted the potential for microreactors to support energy
resilience, remote power applications, industrial process heat, and
national security missions.
A recurring theme in both the stakeholder discussions and the
written correspondence has been the need for the NRC to develop a
clear, predictable, and efficient regulatory framework that supports
rapid licensing of new microreactors and other reactors with comparable
risk profiles and high-volume deployment of these reactors. Several
stakeholders emphasized that when a microreactor applicant demonstrates
low radiological consequences at the site boundary in the unlikely
event of an accident, the NRC should allow the use of a licensing
approach similar to that established for NPUFs. Stakeholders have noted
that such an approach--appropriately adapted for microreactors--would
leverage proven regulatory structures, align safety requirements with
actual risk, and reduce unnecessary regulatory burden while maintaining
the NRC's safety and security standards.

IV. Discussion

A. Need for an Alternative Regulatory Framework

Rapid and high-volume deployment of microreactors and modular
reactors is needed to support national policy and market demand. The
Nuclear Energy Innovation and Modernization Act seeks to streamline
licensing and reduce regulatory uncertainty for advanced reactor
designs. The Accelerating Deployment of Versatile, Advanced Nuclear of
Clean Energy Act requires the NRC to develop ``risk-informed and
performance-based strategies and guidance to license and regulate
microreactors.'' Executive Orders promote the development of domestic
energy supplies to meet the increasing demand for electricity and
direct the NRC to conduct this rulemaking. Market demand for baseload
power has resulted in business cases for high-volume deployment of
microreactors and modular reactors in markets where traditional large-
scale nuclear power plants are impractical or uneconomical.
This proposed rule is needed to establish a regulatory framework
specifically tailored to rapid licensing of first-of-a-kind
microreactors and other reactors with comparable risk profiles and
high-volume deployment of these reactors. The use cases for such
reactors support energy resilience, remote power applications, and
industrial process heat. The proposed framework would be based on
simplified safety requirements and would maximize the benefits of
standardization. The proposed processes and requirements in this rule
would enable shorter licensing timeframes that require fewer resources
than those supported by existing regulations for nuclear power reactors
in part 50 and part 52, which were designed for stationary, large light
water reactors (LWRs). This proposed alternative regulatory framework
is also needed to address Presidential and Congressional direction and
stakeholder feedback.

[[Page 23632]]

B. Description of Proposed Licensing Framework

This proposed rule is complementary to and shares several features
with part 53, ``Risk-Informed, Technology-Inclusive Regulatory
Framework for Commercial Nuclear Plants.'' The part 53 rule features a
risk analysis approach that accommodates licensing all reactor
technologies, including microreactors and large, complex reactors. To
complement this broad scope approach, proposed part 57 would rely on
streamlined safety requirements to focus on simpler license
applications and rapid licensing reviews of new reactors with less
complex designs and operational characteristics and low potential
radiological consequences. The major provisions and features of this
proposed part 57 rule include the following:
1. Rapid Licensing Through Streamlined and Focused Safety Requirements
This proposed rule would provide a pathway to enable rapid
licensing through streamlined and focused safety requirements, for
microreactors and other reactors with comparable risk profiles. The
proposed rule would leverage the simplified designs, limited nuclear
inventory, and overall low risk profiles of these facilities to
establish the necessary and sufficient regulatory requirements to
provide for reasonable assurance of adequate protection. This approach
would enable shorter licensing timeframes by streamlining the
information needed to be prepared by applicants and reviewed by the
NRC. The applicant would be required to submit final design information
and complete operational programs in a joint application for a
construction permit (CP) and associated operating licenses (OLs). The
NRC would conduct a single, comprehensive safety review and potentially
hold one adjudicatory hearing on the joint application. Time and
resource savings would be achieved for qualifying ``first-of-a-kind''
and ``nth-of-a-kind'' designs without any adverse impact on safety and
security.
2. High Volume Licensing
This proposed rule would enable high volume licensing based on
standardization of reactor designs and operational programs. An
applicant would have the option to request a single CP and any number
of OLs for any number of nuclear reactors of essentially the same
design to be built at one or more specific sites or within designated
large geographical areas. Multiple applicants for essentially the same
design would have the option to reference common non-site-specific
information, and the NRC could consolidate some aspects of the
licensing proceedings.
3. Rapid Deployment
This proposed rule would provide options for issuance of a CP to
include approval of the final reactor design and operational programs,
address siting and environmental requirements for large geographical
areas or multiple specific sites, and satisfy requirements for
mandatory and adjudicatory hearings if an applicant provided all
necessary information in a joint application for a CP and associated
OL(s). This could support licensing reactor operation within days of
site selection for time-critical deployment, depending on the
simplicity of onsite construction activities.
4. Multiple Licensing Pathways
The proposed rule would provide several licensing options for
applicants to choose from to meet their deployment model or business
case needs, including a joint application for a CP and associated
OL(s), which would allow for deployment of reactors and approval of
standard designs; a manufacturing license (ML), which would allow for
approval and manufacture of standardized designs and approval of
operational programs; and a standard design approval (SDA), which would
allow for approval of entire reactor designs or major portions thereof.
Applicants would be able to combine requests for these types of
licenses and approvals with requests for license(s), approvals, and
certifications under other regulations in a single application to
holistically address their deployment strategies.
5. Request for Generic Finality
An applicant may include in its joint application for a CP and
associated OL(s) a request for generic finality. Matters resolved in a
proceeding on the application for issuance of the CP and associated
OL(s) for which the applicant has requested and the Commission has
granted generic finality would be considered resolved in proceedings on
other joint applications under proposed part 57 that reference the
approved CP or associated OL(s). For joint applications for ``nth-of-a-
kind'' nuclear reactors and nuclear plants that reference CPs and
associated OL(s) afforded generic finality, the scope of licensing
proceedings would be reduced to site- and applicant-specific
information.
6. Manufacturing License Provisions
The proposed rule would include the use of features to prevent
criticality to allow reactors to be fabricated, fueled, and tested at a
manufacturing facility before being transported to an operating site.
This proposed rule would also allow ML applicants to request and the
NRC to afford finality to the entire nuclear plant design and
operational programs, thereby reducing the scope of proceedings on
joint application for a CP and associated OL(s) that reference the ML
to site- and applicant-specific information.
7. Categorical Exclusions
The proposed rule would permit the use of categorical exclusions
from the requirement for the NRC to prepare an environmental assessment
or environmental impact statement under the National Environmental
Policy Act (NEPA), provided that specific conditions are met.
8. General Licensee for Construction
This proposed rule would establish a general license under which an
applicant that files a joint application for a CP and associated OL(s)
for a ``nth-of-a-kind facility'' could begin construction activities
before the issuance of a CP, provided that certain conditions are met.
9. Alternative to 10 CFR Part 100 Siting Requirements
The proposed rule would allow a graded site characterization
approach with use of existing site characterization data from Federal,
State, or other organizations, provided that the data meets applicable
NRC quality standards.
10. Applicant Defined Definitions
The definitions of many terms in this proposed rule would be
equivalent to the corresponding terms defined in Sec. Sec. 21.3, 50.2,
and 52.1, all entitled ``Definitions,'' and other NRC regulations.
However, given the variety of microreactor and other reactor designs
with comparable risk profiles, flexibility is proposed to allow
applicants to redefine applicable definitions to support their specific
design and licensing basis needs, provided that such redefinitions are
justified and supported by the applicant's safety analysis.
11. Codes or Standards
The proposed rule would allow applicants to propose, with adequate
justification, the use of codes and standards appropriate for their
reactor design and not incorporate by reference

[[Page 23633]]

the specific codes and standards in 10 CFR 50.55a, ``Codes and
standards.''
12. Quality Assurance Program
The proposed rule would not impose quality assurance requirements
under the existing regulations in appendix B, ``Quality Assurance
Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,'' to 10
CFR part 50. Instead, the proposed rule would allow the applicant to
choose an industry-approved quality assurance program, similar to the
approach taken in American National Standards Institute/American
National Standard ANSI/ANS-15.8-1995 (R2018), ``Quality Assurance
Program Requirements for Research Reactors.''
13. Operational Programs
Information related to operational programs concerning facility
operation could be standardized to facilitate fleet-wide deployment of
a microreactor or other reactor with comparable risk profile. These
standardized operational programs could be designed to be administered
onsite or at a corporate or institutional level. Standard operational
programs such as emergency preparedness and security plans would
receive finality, to the extent practicable, for future applicants that
reference those approvals.
14. Remote Monitoring, Remote Operation, and Autonomous Operation
This proposed rule would include provisions for applicants to
specify design features for monitoring and operating a nuclear reactor
from outside the site boundary and for autonomous performance of
operations and safety functions. The NRC has posed a question in this
proposed rule to obtain stakeholder feedback on remote operations and
autonomous operations.
15. Operator Licensing and Human Factors
This proposed rule would adjust staffing, training, personnel
qualifications, and human factors engineering requirements, and would
include provisions for general licenses for reactor operators, to
reflect the expectation that the role of operators would be reduced for
microreactors and other facilities with comparable risk profiles as
compared to the current fleet of large LWRs.
16. Flexible Processes for Changes
This proposed rule includes provisions for ML holders and holders
of OLs that reference reactors manufactured under MLs to combine
applications for license amendments or to make changes to the facility
as described in the final safety analysis report (FSAR) without an
amendment. Under certain conditions, holders of OLs for manufactured
reactors would be able to implement the same changes approved by
amendment to an ML without requesting amendments to their OLs that
reference the ML. This would eliminate duplication of applications for
NRC review of changes to manufactured reactors, including changes that
might be made for improving safety or operational reliability.
17. Readiness for Operation Finding
This proposed rule would provide for the NRC to authorize reactor
operation upon finding that reactor construction conforms to the
approved design and license requirements instead of using inspections,
tests, analyses, and acceptance criteria under 10 CFR part 52, which
could delay this authorization.
18. Fitness-for-Duty Program Flexibility
This proposed rule would allow an applicant to propose an FFD
program of its own specification if operator action would not be
required to maintain the reactor within the criterion of proposed Sec.
57.25(a) or a credible operator or maintenance error could not result
in exceeding that criterion.
19. Resident Inspectors
The NRC does not anticipate stationing a full-time resident
inspector at facilities licensed under this framework. Instead, this
proposed rule would rely on targeted inspections and performance
oversight.
20. Transportation
The proposed rule would add a provision that allows for a risk
methodology to be used for evaluating normal and/or accident conditions
in the event that an applicant cannot meet the testing and performance
requirements of 10 CFR part 71, ``Packaging and Transportation of
Radioactive Material.''
21. Decommissioning and License Termination
The NRC is proposing the flexibility for applicants to develop
decommissioning plans as part of the initial licensing process. This
approach would offer greater flexibility, given the variety of design
and operational strategies being considered. The proposed
decommissioning framework primarily builds on the NPUF model while
incorporating elements from the power reactor framework.
This proposed rule consists of several major components, including
a new part 57, revisions to 10 CFR parts 26, ``Fitness for Duty
Programs,'' and 73, ``Physical Protection of Plants and Materials,''
and conforming changes throughout 10 CFR chapter I to refer to part 57
where appropriate.

C. Utilization Facilities and General Licenses

E.O. 14300 directed the NRC to consider regulating microreactors or
their components through general licenses. Stakeholders also have
expressed interest in the possibility of the NRC using general licenses
for these reactors or redefining ``utilization facility'' to exclude
some nuclear reactors from the licensing requirements in section 103 of
the AEA. The NRC considered these potential alternative approaches for
high-volume licensing and regulation of nuclear reactors or fleets of
reactors in developing this proposed rule. The NRC proposes that using
a general license for regulation of construction activities for certain
structures, systems, and components of nuclear reactors or nuclear
plants would be the most practicable approach under this proposed rule.
The NRC considered whether it would be practicable to exclude
certain reactors that would otherwise be licensed under proposed part
57 from the definition of ``utilization facility'' and regulate them
under a different regulatory framework. The pertinent portions of the
definition of ``utilization facility'' in section 11(cc) of the AEA are
the following: ``(1) any equipment or device, except an atomic weapon,
determined by rule of the Commission to be capable of making use of
special nuclear material in such quantity as to be of significance to
the common defense and security, or in such manner as to affect the
health and safety of the public . . .; or (2) any important component
part especially designed for such equipment or device as determined by
the Commission.'' The AEA definition of a utilization facility allowed
the Atomic Energy Commission (AEC), the NRC's predecessor, to determine
by rulemaking which equipment or devices met the criteria for a
utilization facility. By connecting the definition of a utilization
facility to the quantity of special nuclear material involved and the
manner the material is used, and that material's potential impact on
the common defense and security and public health and safety, Congress
ensured that the AEC's regulatory authority would encompass facilities
whose operation involves radiological safety and security.

[[Page 23634]]

The AEC promulgated a definition of ``utilization facility'' in
1956, now set forth at 10 CFR 50.2 and proposed for part 57, that was
limited to ``any nuclear reactor other than one designed or used
primarily for the formation of plutonium or [uranium-233].'' The AEC
also defined ``nuclear reactor'' as an apparatus, other than an atomic
weapon, designed or used to sustain nuclear fission in a self-
supporting chain reaction. This definition, also part of this proposed
rule, implements both criteria of the AEA's ``utilization facility''
definition. An apparatus designed or used to sustain nuclear fission in
a self-supporting chain reaction meets the first criterion--capable of
making use of special nuclear material (SNM) in such quantity as to be
of significance to the common defense and security. Several current
examples show that even a quantity of SNM less than what is required to
support a self-sustaining fission reaction in a nuclear reactor is
significant to the common defense and security. The U.S. Department of
Energy Order 474.2A, ``Nuclear Material Control and Accountability,''
requires that quantities of uranium-235 or plutonium of 1 gram or
larger are subject to that order and require material control and
accounting and security programs. Additionally, the NRC defines a
quantity of uranium-235 (contained in enriched uranium) in excess of 1
kilogram as being at least Category III material requiring material
control and accounting and security requirements. Finally, the
International Atomic Energy Agency's Nuclear Security Recommendation on
Physical Protection of Nuclear Material and Nuclear Facilities states
that a mass as small as 1 kilogram of uranium-235 (contained in
enriched uranium) needs to be subject to physical security
requirements. These examples are relevant to this proposed rule because
all reactors that would be licensed under this proposed rule--each one
an apparatus designed or used to sustain nuclear fission in a self-
supporting chain reaction--would require more than these minimum
amounts of SNM to operate.
An apparatus designed or used to sustain nuclear fission in a self-
supporting chain reaction also meets the second criterion in the AEA
definition of utilization facility--capable of making use of SNM in
such manner as to affect the health and safety of the public. Decades
of reactor licensing, including research reactors with power levels
ranging from a few watts to several tens of megawatts, have shown that
the use of SNM for self-sustaining fission reactions is capable of
affecting public health and safety. Direct radiation from fission
reactions, the creation and potential release of radioactive
byproducts, and improperly-controlled (or uncontrolled) self-sustaining
fission reactions can all affect public health and safety. Improper
control of a self-sustaining fission reaction can cause significant and
potentially very rapid increases in radiation levels, temperatures, and
pressures, which is why the NRC requires appropriate regulatory
controls that are different than those for devices that use SNM in
other manners, such as a subcritical assembly for physics experiments
or a neutron source for providing the initial neutrons needed to safely
start up a nuclear reactor. These other devices have not typically been
considered utilization facilities. The NRC anticipates that any nuclear
reactor that would be licensed under proposed part 57 to use SNM for
self-sustaining fission reactions for commercial purposes would clearly
require controls to provide reasonable assurance of adequate protection
of public health and safety.
The AEA definition of ``utilization facility'' requires that only
the safety prong or security prong of the definition be met. The
discussion of the safety and security prongs in this document suggests
that any nuclear reactor would meet both prongs and constitute a
utilization facility under the definition in the AEA, thereby
warranting regulation by the NRC as such, consistent with the
responsibilities and authorities conferred to the NRC by the AEA. The
Commission has used its regulatory authority under sections 103 and
182(a) of the AEA to require technical specifications for utilization
facilities to provide reasonable assurance of adequate protection of
public health and safety. The NRC would continue to do so under this
proposed rule.
The NRC considered whether it would be practicable to use the
authority provided to the Commission by section 109(a) of the AEA to
``issue general licenses for domestic activities required to be
licensed under section [101 of the AEA] if the Commission determines in
writing that such general licensing will not constitute an unreasonable
risk to the common defense and security.'' The AEA limits this
authority ``to those utilization and production facilities which are so
determined by the Commission pursuant to section [11(cc)(2)] of [the
AEA].'' Section 11(cc) of the AEA is the definition of utilization
facility, and section 11(cc)(2) of the AEA is ``any important component
part especially designed for [a utilization facility as defined in
section 11(cc)(1) of the AEA] as determined by the Commission.'' Thus,
the NRC can issue a general license for any important component part
especially designed for a utilization facility. The Commission proposes
to use this authority to issue a general license in proposed Sec.
57.45(d) for construction activities, subject to conditions in proposed
Sec. 57.45(d)(1) through (6) that would ensure that the general
license would only be for any important component part especially
designed for a utilization facility, not constitute an unreasonable
risk to the common defense and security, and provide for adequate
protection of the health and safety of the public. The proposed general
license would potentially enable shorter deployment timeframes and is
described in detail in section V.D of this document.
The NRC also considered whether it could include in proposed part
57 a general license for regulation of an entire utilization facility,
meaning a utilization facility as defined in section 11(cc)(1) of the
AEA. However, the AEA provides the NRC with the authority to issue
general licenses only for utilization facilities as defined in section
11(cc)(2) of the AEA, meaning any important component part especially
designed for an entire utilization facility. Therefore, in developing
proposed part 57, the NRC did not consider general licensing of an
entire utilization facility as viable under the current statutory
structure. Instead, the proposed rule would include a licensing
framework under section 103 of the AEA that would reduce the number of
licensing actions, resources for their completion, and required NRC
oversight associated with deployment of individual reactors or nuclear
plants or fleets of such facilities, as described in section IV.B of
this document.

V. Part 57 Framework

A. Discussion of Provisions in Proposed Part 57

Proposed part 57 is comprised of subparts A through Q. These
subparts would provide performance criteria and would be organized to
specify requirements to demonstrate compliance with those performance
criteria throughout the major stages of the life cycle of microreactors
and reactors with comparable risk profiles. The performance-based
approach proposed in part 57 also would include regulatory requirements
that would allow applicants to use a flexible and graded approach to
the performance of

[[Page 23635]]

safety functions based on the role of a particular structure, system,
or component and limiting its impact on assessed radiological
consequence to the public.
Proposed subpart P of part 26 would be new and would be largely
consistent with the fitness-for-duty (FFD) requirements in current
subpart K, ``FFD Programs for Construction,'' of part 26 supplemented
by select requirements from subparts A through I, N, and O of part 26.
These requirements are designed to ensure program effectiveness,
maintain protections afforded to individuals subject to the FFD
program, and align with FFD program implementation by parts 50 and 52
licensees. The proposed requirements would not be entirely equivalent
with requirements in current subpart K of part 26 because the latter
only applies during construction of the nuclear plant, whereas proposed
subpart P of part 26 would apply during construction and operation.
Furthermore, proposed subpart P of part 26 would allow the use of a
variety of biological specimens for drug testing as well as innovative
technologies for drug and alcohol screening and testing that are not
described or allowed by the requirements in subparts A through K, N,
and O of part 26, except under limited conditions.
Proposed part 57 would also include a technology-inclusive
consequence-based approach for physical security and emergency
preparedness for nuclear plants. The NRC used operating experience to
propose additional regulatory flexibility for a part 57 licensee's
implementation of security requirements. This proposed rule would also
propose changes to part 73 for a technology-inclusive approach to
cybersecurity. The proposed provisions for these operational programs
are based on meeting the proposed entry criteria for part 57.
In addition, this proposed rule would make conforming changes
throughout 10 CFR chapter I, by adding ``and part 57'' or similar
language where appropriate to account for the addition of the proposed
part 57.

B. Subpart A--General Provisions

Subpart A would provide the general provisions applicable to all
applicants and licensees under proposed part 57. Subpart A would
include provisions on purpose, scope, definitions, written
communications, deliberate misconduct, employee protections,
completeness and accuracy of information, information collection
requirements, exemptions, standards for review, jurisdictional limits,
attacks and destructive acts, rights related to SNM, license suspension
and rights of recapture, backfitting and issue finality, the Advisory
Committee on Reactors Safeguards, combining licenses, and filing of
applications.
1. Definitions in Proposed Part 57
This proposed rule would provide its own definitions section in
proposed Sec. 57.3, ``Definitions.'' The definitions of many terms in
proposed Sec. 57.3 would be equivalent to the corresponding terms
defined in Sec. Sec. 21.3, 50.2, 52.1, and other NRC regulations.
However, given the variety of microreactor and other reactor designs
with comparable risk profiles, proposed Sec. 57.3 would provide
flexibility by allowing applicants to redefine applicable definitions
to support their specific design and licensing basis needs, provided
that such redefinitions are justified and supported by the applicant's
safety analysis. Definitions established by the application would not
require an exemption from proposed part 57. The flexibility to provide
new definitions would extend only to definitions defined in proposed
part 57 and not to those terms defined by statute, such as ``special
nuclear material.'' Specific proposed definitions are further explained
in the following paragraphs.
The NRC proposes to include a definition of ``Autonomous
operation'' in part 57 that would provide the means for applicants to
present information regarding the performance of operational and safety
functions without reliance on human intervention, external command, or
active control system input under normal operations and accident
conditions. The design of the microreactor with inherent safety
features and active structures, systems, and components (SSCs) would
govern what design functions need to be executed and/or monitored
during normal, off-normal and accident conditions.
The proposed definition of ``Certified fuel handler'' would mean a
non-licensed operator who is responsible for decisions on the safe
conduct of decommissioning activities, safe handling and storage of
spent fuel as defined in 10 CFR 72.3, ``Definitions,'' and appropriate
response to plant emergencies. The certified fuel handler would need to
be qualified in accordance with a fuel handler training program that
meets the same requirements as training programs for non-licensed
operators required by proposed Sec. 57.420, ``Training and
qualification for non-licensed personnel.''
The proposed definition of ``Consensus code or standard'' would be
based on the use of these terms in the National Technology Transfer and
Advancement Act of 1995 (NTTAA) (Pub. L. 104-113) and the Office of
Management and Budget (OMB) Circular No. A-119, ``Federal Participation
in the Development and Use of Voluntary Consensus Standards and in
Conformity Assessment Activities.'' As required by NTTAA, the NRC
undertakes the following activities: (i) consults with voluntary
consensus standards bodies; (ii) participates with voluntary consensus
bodies in the development of consensus standards; and (iii) uses
consensus standards to carry out the NRC's policy objectives.
The proposed definition of ``Construction'' is slightly different
than the current definition in existing Sec. 50.10, ``License
required; limited work authorization.'' The proposed definition would
differ from the current Sec. 50.10 definition in that it would apply
to only safety-related SSCs (as defined in proposed part 57) and SSCs
relied upon to implement the proposed security requirements.
The proposed definition of ``Control room'' would provide a means
for remote monitoring and/or remote operation outside the site boundary
where actions can be taken to operate the nuclear power unit safely
under normal conditions and to maintain it in a safe condition under
accident conditions.
The proposed definition of ``Decommission'' would be slightly
different than the definition in Sec. 50.2. The proposed definition
would also include permanent removal of an individually licensed
nuclear reactor.
The proposed definition of ``Defense in depth'' would provide a
philosophy of designing a nuclear facility that includes two or more
independent and redundant layers of defense in the design of a facility
and its operating procedures to compensate for uncertainties such that
no single layer of defense, no matter how robust, is exclusively relied
upon. Defense in depth includes, but is not limited to, the use of
access controls, physical barriers, redundant and diverse safety
functions, and emergency response measures.
The proposed definition of ``Design bases'' would be the
information that identifies the specific functions to be performed by
an SSC of a facility, and the specific values or ranges of values
chosen for controlling parameters as reference bounds for design. These
values may be (1) restraints derived from generally accepted ``state-
of-the-art'' practices for achieving functional

[[Page 23636]]

goals, or (2) requirements derived from analysis (based on calculation
and/or experiments) of the effects of a postulated accident for which
an SSC must meet its functional goals.
The proposed definition of ``Design features'' would be the active
and passive SSCs and inherent characteristics of those SSCs that
contribute to limiting the total effective dose equivalent (TEDE) to
individual members of the public during normal operations and prevent
or mitigate the consequences of design basis accidents.
The proposed definition of ``Fission product release'' would be the
amount and composition of radioactive material released to the
environment, after accounting for any retention of radionuclides
provided by reactor design features.
The proposed definition of ``Fuel'' would be SNM or source
material, discrete elements that physically contain SNM or source
material, and homogeneous mixtures that contain SNM or source material,
intended to or used to create power in a nuclear reactor.
The proposed definition of ``Licensing basis information'' would be
the information contained in regulations, orders, licenses,
certifications, or approvals issued by the NRC for a nuclear plant
licensed under proposed part 57 and that information submitted to the
NRC by an applicant or licensee in a safety analysis report, program
description, or other licensing-related document required under
proposed part 57.
The proposed definition of ``Manufactured reactor'' would be the
essential portions of a nuclear reactor that are manufactured under an
ML and subsequently incorporated into a nuclear plant under a
construction permit issued under subpart C of proposed part 57.
The proposed definition of ``Manufacturing license'' would be a
license issued under subpart D of proposed part 57 that authorizes the
production of manufactured reactors but not their construction,
installation, or operation.
The proposed definition of ``Programmatic controls and operational
programs'' would be administrative procedures that govern human action
in implementing programs and operating, monitoring, and maintaining
SSCs and equipment of a nuclear plant. Programmatic controls could be
standardized to facilitate fleet-wide deployment of a microreactor.
These standardized operational programs could be designed to be
administered on site or at a corporate or institutional level.
Implementation milestones for each operational program would need to be
described depending on whether the program will be implemented all at
once or on a phased basis.
The proposed definition of ``Quality assurance'' (QA) would be
planned and systematic actions during design, construction, and
modification necessary to provide adequate confidence that the SSC will
perform satisfactorily in service.
The proposed definition of ``Remote monitoring'' would mean
observing plant data from a location outside of the site boundary.
Remote monitoring does not include the performance of any operator
actions necessary to manipulate the reactor to protect the public
health and safety (i.e., remote operations). However, remote monitoring
could be used to access real-time data needed to perform other
functions that protect the public health and safety, such as emergency
preparedness or security. The ability to protect the public would be
dependent upon having accurate and timely access to the plant-monitored
parameter data. Wireless communication could be used to support remote
monitoring.
The proposed definition of ``Remote operation'' would be to command
and control the reactor from a location outside of the site boundary.
Industry has indicated that the design of a microreactor with inherent
safety features and active SSCs would govern what design functions need
to be executed and/or monitored during normal, off-normal, and accident
conditions.
The proposed definition of ``Safe shutdown'' would be bringing the
nuclear reactor to safe, stable conditions specified in plant technical
specifications when the reactor is under design basis accident
conditions with loss of emergency power and offsite power.
The proposed definition of ``Safety function'' would be the purpose
served by a design feature, human action, or programmatic control to
prevent or mitigate unplanned events and thereby demonstrate compliance
with requirements in proposed part 57 for limiting risks to public
health and safety. Safety functions could be performed by any
combination of the elements supported by the safety analysis and could
be specified at the plant level or at the level of a particular barrier
or system. Multiple plant-level safety functions would be assumed to
apply to all reactor designs based on established requirements and
historical practices. These fundamental safety functions would include
the control of reactivity, removal of heat, and limiting the release of
radioactive materials. The protection of a specific barrier or system
that contributes to meeting plant-level safety criteria could also be
referred to as a safety function.
The proposed definition of ``Safety-related structures, systems and
components'' is slightly different than the definition in Sec. 50.2.
Whereas the Sec. 50.2 definition refers to ``events,'' the proposed
definition would refer to ``accidents.'' Design basis accidents bound
events. Also, where the Sec. 50.2 definition refers to a reactor
coolant pressure boundary, the proposed definition would be technology
neutral because some reactor designs under proposed part 57 may not
operate at pressure.
The proposed definition of ``Source term'' would be 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. The
source term would be developed by the applicant when performing the
maximum hypothetical accident (MHA) or maximum credible accident (MCA)
methodology. This source term would then be analyzed with site
parameter information to demonstrate compliance with the accident dose-
based entry criterion in proposed Sec. 57.25(a).
The proposed definition of ``Special nuclear material'' would be
(1) plutonium, uranium-233, uranium enriched in the isotope-233 or in
the isotope-235, and any other material that the Commission, pursuant
to the provisions of section 51 of the AEA, determines to be SNM, but
does not include source material; or (2) any material artificially
enriched by any of the foregoing, but does not include source material.
2. Other General Provisions
Proposed Sec. 57.4, ``Written communications,'' would govern
written communications and how applications and other required
information must be submitted to the NRC. These requirements would be
equivalent to those in Sec. 50.4, ``Written communications.''
Proposed Sec. 57.5, ``Deliberate misconduct,'' would establish
requirements for enforcement action to which a licensee, an applicant,
or a licensee's or applicant's contractor or subcontractor, or an
employee of any of them, may be subject for engaging in deliberate
misconduct. These requirements would be equivalent to those in Sec.
50.5, ``Deliberate misconduct.''

[[Page 23637]]

Proposed Sec. 57.6, ``Employee protection,'' would prohibit
discrimination against an employee of a holder or applicant for an NRC
license, permit, or SDA, or a contractor or subcontractor of a holder
or applicant for an NRC license, permit, or SDA for engaging in certain
protected activities. Proposed Sec. 57.6 also would prescribe a
procedure for seeking a remedy for employees who believe they have been
discriminated against for engaging in such protected activities. These
requirements would be equivalent to those in Sec. Sec. 50.7 and 52.5,
both entitled ``Employee protection.''
Proposed Sec. 57.7, ``Completeness and accuracy of information,''
would govern the completeness and accuracy of information provided to
the NRC. These requirements would be equivalent to those in Sec. Sec.
50.9 and 52.6, both entitled ``Completeness and accuracy of
information.''
Proposed Sec. 57.8, ``Information collection requirements: OMB
approval,'' would establish requirements for information collection
requirements and OMB approval. These requirements would be equivalent
to those in Sec. 50.8, ``Information collection requirements: OMB
approval.''
Proposed Sec. 57.9, ``Specific exemptions,'' would govern
exemptions from the requirements of the regulations in proposed part
57. These requirements would be equivalent to those in Sec. Sec. 50.12
and 52.7, both entitled ``Specific exemptions.''
Proposed Sec. 57.11, ``Jurisdictional limits,'' would require that
no license or SDA issued under proposed part 57 would cover activities
that are not under or within the jurisdiction of the United States.
These requirements would be equivalent to those in Sec. 50.53,
``Jurisdictional limitations.''
Proposed Sec. 57.12, ``Attacks and destructive acts,'' would state
that licensees, holders of standard design approvals, and applicants
for licenses and standard design approvals would not be required to
provide design features or other measures for the specific purpose of
protection against the effects of attacks and destructive acts by
enemies of the United States directed against the facility or
deployment of weapons incident to U.S. defense activities. These
requirements would be equivalent to those in Sec. 50.13, ``Attacks and
destructive acts by enemies of the United States; and defense
activities.''
Proposed Sec. 57.13, ``Rights related to special nuclear
material,'' would establish requirements for rights related to SNM.
These requirements would be equivalent to those in Sec. 50.54(b) and
(c).
Proposed Sec. 57.14, ``License suspension and rights of
recapture,'' would establish requirements for license suspension and
rights of recapture of the material or control of the facility in a
state of war or national emergency declared by Congress. These
requirements would be equivalent to those in Sec. 50.54(d).
Proposed Sec. 57.15, ``Agreement limiting access to Classified
Information,'' would address requirements for agreements limiting
access to classified information and would be equivalent to Sec.
50.37, ``Agreement limiting access to Classified Information.''
Proposed Sec. 57.16, ``Backfitting and issue finality,'' would
address backfitting requirements by providing requirements that would
be equivalent to those in Sec. 50.109, ``Backfitting,'' and issue
finality requirements by providing requirements that would be
equivalent to those in Sec. Sec. 52.83(a), 52.145, ``Finality of
standard design approvals; information requests,'' and 52.171,
``Finality of manufacturing licenses; information requests.'' An
exception is that proposed Sec. 57.16(c) would not include an
equivalent requirement to Sec. 52.171(b)(2), which requires the
Commission to determine that departures will comply with the
requirements in Sec. 52.7 and that the special circumstances for the
departure would outweigh any decrease in safety that may result from
the reduction in standardization caused by the departure. Proposed
Sec. 57.16(c) would instead require the joint application for the
referencing CP and OL(s) to include analysis of departures from the
design characteristics, site parameters, terms and conditions, or
approved design of the nuclear reactor, nuclear plant, or manufactured
reactor. Proposed Sec. 57.16(c) would also specify that analysis would
not be required for departures from any operational programs or
requirements approved with the referenced CP, OL, or ML that are not
material to the adequacy of the design, if the joint application
includes proposed alternative operational programs or requirements.
Under proposed Sec. 57.16(c), all departures would be subject to
litigation in the same manner as other issues in the CP or OL, which
would be equivalent to Sec. 52.171(b)(2).
Proposed Sec. 57.17, ``Referral to the Advisory Committee on
Reactor Safeguards (ACRS),'' would address referral to the Advisory
Committee on Reactor Safeguards (ACRS) and would be equivalent to
Sec. Sec. 50.58, ``Hearings and report of the Advisory Committee on
Reactor Safeguards,'' 52.141, ``Referral to the Advisory Committee on
Reactor Safeguards (ACRS),'' and 52.165, ``Referral to the Advisory
Committee on Reactor Safeguards (ACRS).''
Proposed Sec. 57.18, ``Combining licenses; elimination of
repetition; relationships between subparts,'' would address combining
applications and would be equivalent to Sec. Sec. 50.31, ``Combining
applications,'' 50.52, ``Combining licenses,'' and 52.8, ``Combining
licenses; elimination of repetition.'' Proposed Sec. 57.18 would also
provide clarity about various combinations of licenses and contents of
related applications that would enable various high-volume deployment
strategies. While proposed part 57 clearly outlines the licensing
framework for combining licenses for multiple reactors, multiple sites,
manufacturing, possession of special nuclear material, and other
deployment activities, this licensing framework largely exists under
other parts of 10 CFR chapter I, such as parts 50, 52, and 53.
Proposed Sec. 57.18(a)(1) would include a provision for
applications that would be filed under proposed part 57 by one or more
applicants for licenses to construct and operate nuclear reactors or
nuclear plants of essentially the same design to be located at
different sites, to refer to a single FSAR. This proposed provision
would be similar to the provisions in appendix N to part 50,
``Standardization of Nuclear Power Plant Designs: Permits To Construct
and Licenses To Operate Nuclear Power Reactors of Identical Design at
Multiple Sites.''
Proposed Sec. 57.18(a)(2) would include a provision that an
applicant may include in one application for a CP and associated OL(s)
for a nuclear reactor or nuclear plant under proposed part 57
information for multiple sites at which the applicant proposes to
construct and operate the reactor or plant. This proposed provision
would allow for licensing construction and operation of a single
nuclear reactor or nuclear plant at multiple locations over its
lifetime, such as for operational testing at a manufacturing facility
and power operation at a deployment site.
Proposed Sec. 57.18(a)(3) would require an application under
proposed part 57 for multiple types of permits, licenses, or
certifications to clearly indicate to which permit, license, or
certification information in the application pertains. This proposed
requirement would facilitate the NRC's review of the application by
ensuring that the NRC would apply the appropriate proposed requirements
(e.g., standards of review, issuance, hearings, finality, etc.) to the
information in the application.

[[Page 23638]]

Proposed Sec. 57.18(a)(4) would include provisions for holders of
OLs that reference the same ML to combine among themselves, or with the
holder of the ML, applications for license amendments under proposed
Sec. 57.310, ``Amendment of license.'' This proposed provision would
potentially decrease the overall resources that would be required for
applicants and the NRC for identical requests for amendments to
multiple licenses as opposed to separate filings and reviews of each
application for amendment.
Proposed Sec. 57.18(a)(5) would specify that an applicant may
include in a single joint application a request for a CP for any number
of nuclear reactors of essentially the same design that would be built
at a specific site and requests for OLs for those reactors, provided
that the application would state the earliest and latest dates for
completion of the construction of each nuclear reactor as would be
required by proposed Sec. 57.55(g) and would include the information
that would be specified in proposed Sec. 57.60(a)(4). This proposed
provision would potentially reduce applicant and NRC resources related
to licensing a nuclear plant at which multiple nuclear reactors of
essentially the same design would be operated over its lifetime,
including replacement reactors.
Proposed Sec. 57.18(b), (d), and (e) would include provisions for
incorporating by reference information contained in previous
applications, statements, or reports filed with the Commission and
applicable Commission approvals issued under part 50 or 52; referencing
a standard design approval, CP, OL, ML, or combination thereof, that
would be issued under proposed part 57; and referencing a relevant U.S.
Department of War or U.S. Department of Energy authorization for a
utilization facility that has been tested and that has demonstrated the
ability to function safely, respectively. These provisions would allow
applicants and the NRC to minimize duplication of previous efforts in
filing and reviewing applications under proposed part 57.
Proposed Sec. 57.18(c) would continue the Commission's practice of
combining multiple authorizations for a licensee under various parts of
10 CFR chapter I into one license based on the Commission's authority
under section 161(h) of the AEA to combine NRC licenses.
Proposed Sec. 57.19, ``Filing of application,'' would address
filing of applications and would be equivalent to Sec. Sec. 50.30,
``Filing of application; oath or affirmation,'' 52.135, ``Filing of
applications,'' and 52.155(a). Proposed Sec. 57.19(f) would require an
applicant for licenses to construct and operate one or more nuclear
reactors under subpart C of proposed part 57 to file a joint
application for a CP and associated OL(s). Proposed Sec. 57.19(f)
would also require that the joint application include the information
specified in proposed Sec. Sec. 57.55, ``Content of applications;
general information,'' and 57.60, ``Content of applications; technical
information,'' and be complete enough to permit all evaluations
necessary for the issuance of the requested CP and the associated OL(s)
upon the NRC making the finding required by proposed Sec. 57.100(b)(1)
(i.e., the finding that construction has been substantially completed).
The joint application would permit the NRC to use the regulations in
Sec. 2.105(c) to specify in the notice of proposed issuance of the CP
that on completion of construction and the NRC making the finding that
would be required by proposed Sec. 57.100(b)(1), the associated OL(s)
would be issued without further prior notice, thus streamlining the
process for issuance of the associated OL(s) and reducing the timeframe
for licensing.

C. Subpart B--Eligibility

The NRC based the development of the proposed part 57 framework on
existing licensing practices for non-power and other utilization
facilities that, by design and operational characteristics, present low
risks of radiological consequences. These characteristics have
designers approach safety by emphasizing accident prevention with
inherent self-limiting reactivity feedback mechanisms and passive
safety systems for heat and decay heat removal without reliance on
complex active safety systems. The NRC used these characteristics to
create a set of requirements to determine which applicants would be
eligible to use proposed part 57. Located in proposed Sec. Sec. 57.25,
``Applicability,'' and 57.30, ``Design criteria attributes,'' these
proposed requirements are termed ``entry criteria'' and ``design
criteria attributes,'' respectively.
Given the wide range of reactor types and their functional
characteristics, this proposed rule would emphasize the ``attributes''
of microreactors and other reactors with comparable risk profiles.
Rather than defining these reactors in terms of thermal power level,
this attribute-based approach would describe microreactors and other
reactors with comparable risk profiles in terms of their functional
characteristics, such as the capability to prevent or mitigate
accidents without active systems or operator intervention. By doing so,
the NRC recognizes that reactors with inherently safe design features
and more favorable safety profiles may appropriately be designed with
higher power levels than other reactor designs.
The first eligibility criterion would be a dose-based acceptance
value. The second eligibility criterion would be an upper limit on the
amount of fuel. These eligibility criteria are intended to screen in
reactor designs that are smaller, simpler, and more conducive to rapid,
high-volume licensing. These eligibility criteria would be supported by
six design criteria attributes. These design criteria attributes
emphasize the features of inherently and passively safe reactors that
make them secure and protective against radiological harm. These
attributes include (1) reactivity control, (2) heat removal, (3)
fission product retention, (4) shielding, (5) radioactive effluents
control, (6) security by design. If an applicant for a reactor design
does not meet these criteria, they can apply for a license under a
different regulatory framework.
1. Dose-Based Entry Criterion
A dose-based entry criterion under accident conditions would be
used to inform the analysis of postulated accidents and the development
of safety measures so that, in the unlikely event of an accident, there
is assurance that no acute radiation-related harm will result to any
member of the public. The Commission has found the use of a dose-based
entry criterion to be adequate for facility siting and design purposes
based on decades of extensive experience in the criterion's application
and in recognition of the assumptions and considerations applied within
the radiological consequence analyses. While the dose-based entry
criterion would be computed in terms of dose, it is a figure of merit
used to characterize the minimum requirements for design, fabrication,
construction, testing, operational limits, and performance for safety-
related SSCs. The numerical value of the criterion does not represent
acceptable or actual public exposures received during normal and
emergency conditions, which are primarily controlled by 10 CFR part 20,
``Standards for Protection Against Radiation,'' and through emergency
planning.
An applicant would be required to demonstrate their reactor design
meets the 1 rem (10 millisieverts (mSv)) TEDE dose-based entry
criterion in proposed Sec. 57.25(a), and the NRC has found that the
maximum hypothetical and

[[Page 23639]]

maximum credible accident methodologies would be acceptable means of
providing this demonstration. These methodologies are associated with a
fission product release accompanying damage to fission product
retention barriers, maximum allowable leak rates, a postulated single
failure of any safety-related SSCs, conservative site meteorological
dispersion characteristics, and an individual member of the public
presumed to be at the location of maximum cumulative dose in the
unrestricted area without protective actions. By demonstrating under
these conservative assumptions that, in the unlikely event of an
accident, the dose to the maximally exposed individual member of the
public in the unrestricted area would remain below the accident dose
acceptance criterion, there is reasonable assurance that actual
accidents would not result in acute offsite doses.
Historically, NRC licensing processes have relied on deterministic
bounding analyses that, while conservative, may impose unnecessary
siting, design, and operational constraints on advanced reactor designs
with inherent and highly reliable passively safe reactor technologies.
The Commission recognizes the need for flexibility in how applicants
define their licensing basis to reflect the diversity of microreactors
and other reactor designs with comparable risk profiles. Proposed part
57's inclusion of both the MHA and MCA methodologies provides risk-
informed and performance-based regulatory pathways that align the
applicant's safety analysis scope with the complexity and safety
characteristics of their design. Proposed part 57 distinguishes between
the MHA and the MCA with respect to the amount of analytical rigor
necessary to justify the derived source term. By distinguishing between
the MHA and MCA approaches, the Commission would allow applicants to
tailor the scope and depth of their accident analyses to their design
and business model needs while continuing to ensure safety.
The source term defines 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. The applicant would utilize their MHA or
MCA source term to establish the site boundary and determine the level
of design, qualification, testing, and maintenance of SSCs necessary to
show with reasonable assurance that the radiological consequences at
the site boundary are below the 1 rem TEDE entry criterion of proposed
Sec. 57.25(a).
Depending on the desired level of analysis, applicants may select
either the MHA or MCA approach. The MHA approach can demonstrate safety
through a postulated accident scenario, often highly conservative,
which assumes a severe release of radioactive material consistent with
physical laws, regardless of probability. This MHA analysis does not
rely on detailed risk-informed assessment methodologies, thereby
reducing analytical complexity for reactors with few to no active
systems or self-limiting physical phenomena. The MHA approach may be
desirable for applicants that are willing to accept additional
conservatism by leveraging simplified analyses that are less time and
resource intensive. Although the MHA may not necessarily reflect a
realistic or credible sequence of events, it represents a bounding case
to support subsequent safety decisions.
If an applicant does not wish to accept the conservatisms
associated with the MHA approach, further analyses would need to be
performed to support an MCA approach. The MCA approach excludes certain
physically unrealistic or excessively conservative assumptions,
focusing instead on events that are credible given the technology,
safety systems, and plant operating conditions. The MCA analysis can
leverage a variety of modern risk-informed methodologies to credibly
quantify events and consequences, providing a rational basis for a
smaller site boundary and focused SSC categorization and potentially
reducing the number of components subject to the more stringent safety
requirements.
Two identical reactor designs could, in principle, yield different
site boundary distances and safety classifications depending on whether
their analyses employ the MHA or MCA methodology. Under the MHA
approach, conservative bounding assumptions, such as postulated worst-
case system failures and maximum radionuclide release, would produce a
larger source term necessitating a greater site boundary and broader
safety classification of SSCs. In contrast, an MCA analysis that
quantifies system performance and reliability could justify a smaller,
more realistic source term and a correspondingly smaller site boundary
and narrower safety classification. Both outcomes would be acceptable
under proposed part 57's consequence-based framework because each would
provide reasonable assurance that offsite radiological consequences
remain below the 1 rem TEDE entry criterion. The preferred approach
would likely depend on the scope and depth of analysis the applicant
wishes to undertake. Applicants would need to be clear on which
approach is being applied, and analyses would have to be supported by
appropriate and sufficient technical justifications.
The NRC is providing flexibility on how the TEDE dose-based entry
criterion would be met in recognition of the need for expedited
licensing and deployment of the types of facilities on which proposed
part 57 is focused. Including both the MHA and MCA methodologies
supports the Commission's regulatory modernization goals by encouraging
innovation in reactor design while maintaining a consistent safety
objective. Furthermore, this graded approach would enable efficient
licensing reviews by aligning analytical rigor with risk significance
without diminishing safety assurance. Under this proposed framework,
applicants should discuss their plans for use of an MHA or MCA with the
NRC staff prior to submittal of an application. This would ensure there
is common understanding of the applicant's approach and would allow for
resolution of any issues before development of a complete application.
2. Fuel Mass Limit
The premise of this proposed rule is to establish regulatory
requirements commensurate with the low hazards posed by facilities that
would be licensed under proposed part 57. These requirements would be
justified by the use of a dose-based entry criterion applied to the
results of a maximum hypothetical or maximum credible accident that
assesses siting and the performance of safety-related SSCs. This would
also be true for large LWRs with a very large site boundary. However,
many of the traditional requirements that the NRC considered when
creating this proposed rule have historically provided defense in depth
to address unlikely events that may exceed analyzed releases.
Traditional requirements include the Commission's historical treatment
of severe accidents based on lessons learned from operating large LWRs.
Examples of these regulations include: 10 CFR 50.46, ``Acceptance
criteria for emergency core cooling systems for light-water nuclear
power reactors,'' for assessing large-break loss of coolant accidents;
10 CFR 50.155, ``Mitigation of beyond-design-basis events,'' for
flexible mitigation strategies for beyond-design-basis events; and
several part 52 requirements for severe accident design features.

[[Page 23640]]

The fuel mass limit entry criteria would deterministically screen
reactor designs without additional performance-based acceptance
criterion or severe accident analysis to assess events beyond which
SSCs could be challenged. The fuel mass limit entry criteria would be
established to provide additional defense in depth for these very
unlikely events by limiting the amount of decay heat that may
necessitate the need for active cooling systems and overall material
available for release, further limiting the potential for causing acute
health effects to the public. However, the NRC has proposed a question
in this proposed rule, asking whether, in lieu of applying a
deterministic material limit on the quantity of SNM, the NRC should
apply an alternative performance-based acceptance criterion such as an
adiabatic heat rate threshold, beyond which SSCs could be challenged.
To assist in developing a quantitative basis for such a limit, the
NRC reviewed and evaluated the quantities of SNM in the cores of
several reactor types. In evaluating the quantities of SNM, the NRC
determined the quantities of uranium (U) and plutonium (Pu). This
includes the following isotopes: \1\ U-233, U-234, U-235, U-236, U-238,
Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Pu-244. For
technological neutrality, the mass criteria would also include thorium
isotopes, because thorium can be used as a breeding material in thermal
spectrum breeder reactors. None of the reactors considered in the
evaluation included this technology, but there have been early
indications of industry interest in pursuing this concept.
---------------------------------------------------------------------------

\1\ None of the evaluated non-LWRs included thorium, so they had
negligible amounts of U-233.
---------------------------------------------------------------------------

In conducting this evaluation, the NRC considered a spectrum of
reactor technologies, including several non-LWR designs, two small
modular pressurized water reactors (PWRs) and one small modular boiling
water reactor (BWR), and several representative large LWRs. The purpose
of this evaluation was to understand the similarities and differences
between these reactor technologies and inform an entry criterion that
facilitates high-volume licensing of microreactors. The assessment
compared these reactor technologies, the SNM masses, type and kinds of
engineered safety features, and accident response characteristics. To
perform this evaluation, the NRC considered several sources of publicly
available information covering a range of reactor types and power
levels.
The evaluation included several non-LWRs of various reactor types
and fuel forms (e.g., TRISO, metal, oxide, and molten salt) and
coolants (e.g., gas, molten salt, liquid metal, water). The power range
of these designs spans from approximately 5 megawatts thermal
(MW<INF>th</INF>) to about 2250 MW<INF>th</INF>. The assessment also
included small modular and large LWRs to gain a sense of the
differences in SNM quantities between the non-LWR and small LWR designs
currently in development versus the quantities in the currently
operating large LWR commercial fleet. The power reactor range for the
large LWRs spans from approximately 2600 MW<INF>th</INF> to about 4400
MW<INF>th</INF>.
The quantities of SNM vary by reactor technology. For each reactor
technology, the NRC calculated SNM quantities at the beginning and end
of an operating cycle based on published core and fuel parameters and
operational characteristics. To perform the calculation, the NRC
utilized the Oak Ridge National Laboratory SCALE code system. The SCALE
code system is a widely used modeling and simulation suite for nuclear
safety analysis and design. Results of these calculations found that
the large LWR SNM quantities at the beginning of an operating cycle
ranged from approximately 71 metric tons heavy metal (MTHM) \2\ for a
PWR to 154 MTHM for a BWR. At the end of an operating cycle, these
quantities range from approximately 69 to 148 MTHM, respectively.
Except for a large molten salt reactor, which had an SNM quantity of
approximately 43 MTHM, the remaining reactors at the beginning of an
operating cycle had SNM quantities no greater than 9.3 MTHM and at the
end of an operating cycle, or equilibrium, SNM quantities no greater
than 8.7 MTHM.
---------------------------------------------------------------------------

\2\ MTHM is a unit used to define the mass of SNM where that
material may include more than uranium (i.e., when plutonium is
included). One metric ton of heavy metal equates to 1000 kg of
uranium, plutonium, or both. For a reactor containing entirely
uranium fuel, 1 MTHM = 1 MTU.
---------------------------------------------------------------------------

Table 1 compares various reactor types by the amount of SNM, in
terms of MTHM, each contains by cycle period. Table 1 provides the
reactor name, fuel type, percent fuel enrichment, and cycle period for
which each of the SNM quantities were estimated as beginning of life
(BOL), continuous refueling (cont.), equilibrium (equil.), beginning of
equilibrium cycle (BOEC), and end of equilibrium cycle (EOEC). The BOL
are conditions of the reactor core at initial startup after fresh fuel
loading. The end of life (EOL) describes the conditions of the reactor
core at the end of its useful fuel cycle, when fuel burnup or
reactivity limits have been reached. Some reactor designs operate
continuously. For continually refueled systems, SNM inventories are
given as equilibrium conditions. For these designs, the BOEC is a state
of the reactor core at the start of a cycle once equilibrium operating
conditions have been established. Likewise, the EOEC is a state of the
reactor core operating on a continuous refueling cycle at the end of a
typical equilibrium operating cycle, after equilibrium burnup has
occurred. Uranium dioxide (UO<INF>2</INF>) is a ceramic oxide fuel made
from uranium dioxide powder, pressed into pellets, and sintered for
LWRs. TRISO fuel consists of spherical uranium kernels, usually of
uranium dioxide or uranium oxycarbide, coated with multiple layers of
pyrolytic carbon and silicon carbide, which act as a miniature
containment system. Metallic alloy fuel in a compact form is composed
of uranium (U), transuranics (TRU), and 10 weight percent (wt. %)
zirconium (Zr) (U-TRU-10Zr Metal Fuel). Molten salt fuel is a liquid
fuel salt mixture consisting of lithium fluoride (LiF), beryllium
fluoride (BeF<INF>2</INF>), and uranium tetrafluoride (UF<INF>4</INF>)
(LiF-BeF<INF>2</INF>-UF<INF>4</INF>).
BILLING CODE 7590-01-P

[[Page 23641]]

[GRAPHIC] [TIFF OMITTED] TP01MY26.006

BILLING CODE 7590-01-C
Reactor safety profiles vary significantly between technologies due
to differences in fuel type, coolant, operating characteristics, and
reliance

[[Page 23642]]

on active versus intrinsic and passive safety systems. Traditional
large LWRs have large inventories of SNM and operate at higher power
levels, power densities, and operating pressures than the other
reactors studied. These features present more complex accident
scenarios, and the reactor design relies on multiple engineered safety
systems, active cooling, and robust containment structures to manage
accident conditions. Accident analyses for large LWRs frequently
require a high level of analytical rigor, including the use of
sophisticated probabilistic risk assessment methodologies and
computational tools to characterize plant responses and overall risk
profiles. While appropriate for complex, high-power facilities, this
level of analysis is resource intensive and not well suited to the
streamlined processes needed to support high-volume licensing. In
contrast, many advanced non-LWR designs incorporate inherent safety
features--such as low-pressure operation, high thermal capacities, and
strong negative reactivity feedbacks--that reduce the likelihood and
severity of accidents. Also, small LWRs, while similar in technology to
large LWRs, generally benefit from reduced core power levels and power
density, fission product inventories, and simpler system layouts,
leading to more straightforward accident analyses. As such, these non-
LWR and small LWR risk profiles can demonstrate the designs' low
consequence without a very large site boundary and without extensive
reliance on probabilistic risk assessment methods. These safety
features and relatively small sizes and source terms as compared to
large LWRs lend themselves to licensing and manufacturing
standardization, which makes these types of reactors more conducive to
efficient, high-volume licensing.
To understand the various reactor technology safety profiles, the
NRC reviewed several published scientific studies, NRC's preliminary
safety evaluation reports, and environmental review documents. The
review focused on identifying common design attributes among these
reactors--such as strong negativity reactivity feedback, robust fuel
forms, higher thermal margins, and passive heat removal--that
inherently limit transient and accident progression. The NRC found non-
LWR designs and microreactors are often designed with large thermal
capacities that allow them to dissipate operational and decay heat
passively for relatively long periods of time without the need for
active systems or operator action. These designs also feature large
shutdown reactivity margins and other intrinsic safety characteristics
that provide strong inherent barriers to accident progression. As a
result, their overall safety behavior can be well understood without
relying on sophisticated probabilistic or risk assessment
methodologies, since the fundamental design attributes themselves
demonstrate a robust ability to prevent and mitigate accidents that
previous large LWR designs have traditionally been designed to
accommodate. Accordingly, these designs do not necessarily have the
need for traditional containments as there is a reduced likelihood of
events occurring requiring such mitigation features. Furthermore, these
designs would not warrant precautionary protective measures to respond
to emergencies. Instead, as a final layer of defense in depth,
licensees could rely on a risk-informed approach to emergency planning.
Based on its evaluation of SNM inventories and safety
characteristics of non-LWRs, small LWRs, and representative large LWRs,
the NRC concluded that the establishment of a defined SNM material
limit would be technically justified as an entry criterion to proposed
part 57. This material limit would be defined as a total inventory of
thorium, uranium, and plutonium contained in the nuclear reactor not to
exceed 10 metric tons. The evaluation showed that designs within the
material limit would likely have inherent and passive safety features
and exhibit favorable safety profiles despite variations in core design
and thermal power levels. Together, these insights support the NRC's
determination that a numerical material limit that is risk-informed due
to inherent and passive design features could be part of an appropriate
regulatory threshold to using a licensing approach to enable rapid and
efficient licensing of microreactors and other reactor designs with
comparable risk profiles.
3. Design Criteria Attributes
The design criteria attributes in proposed Sec. 57.30--reactivity
control, heat removal, fission product retention, shielding,
radioactive effluent control, and security by design--are rooted in the
fundamental principles of nuclear safety and radiation protection.
<bullet> Reactivity Control--The reactor would need to be able to
safely control the power level in normal operation, shut down quickly
if needed, and stay safely shut down. The reactor would be required to
have a natural ``braking'' effect: when temperatures rise, the power
level automatically falls (net negative reactivity feedback). Also, if
the fuel would be loaded into the reactor at a manufacturing facility,
then the reactor design would need to have built-in protections to
prevent the reactor from unplanned criticality.
<bullet> Heat Removal--Even after the reactor is shut down, heat
keeps being produced. The design would be required to have highly
reliable, passive systems to keep the reactor cool and within safe
temperature limits, even if the main cooling system fails during events
like power loss or earthquakes.
<bullet> Fission Product Retention--Barriers like the fuel itself
and the reactor vessel can retain radioactive materials during both
normal operations and accident conditions. The design would need to
keep temperatures and pressures well below the limits these barriers
can handle.
<bullet> Shielding--The reactor would need strong, durable
shielding to protect workers and the public from radiation, including
during transportation. The design also would have to account for heat
that builds up in shielding and the removal of the heat if needed.
<bullet> Radioactive Effluents Control--The reactor would be
required to meet limits for any radioactive gases, liquids, or solid
wastes it would release, and have monitoring and handling systems that
protect people and the environment.
<bullet> Security by Design--Where possible, the design itself
should address security risks, using built-in engineering and physical
protection features instead of relying only on procedural measures.

D. Subpart C--Construction Permits and Operating Licenses

Proposed subpart C would provide requirements related to
applications for NRC licenses to construct and operate utilization
facilities for commercial or industrial purposes under part 57. The AEA
calls these licenses ``construction permits'' and ``operating
licenses,'' and the NRC proposes to use that nomenclature in proposed
part 57 as it has done in part 50. Proposed part 57 would include
licensing options based on the CP and OL approaches in part 50, and
proposed subpart C would contain several sections that would be similar
to existing regulations in part 50.
Proposed Sec. 57.45, ``License required; exceptions from
licensing,'' would address required licenses and identify certain
exceptions from licensing. Proposed Sec. 57.45(a) would describe
activities requiring an NRC license and would be equivalent to Sec.
50.10(b). Proposed Sec. 57.45(b) would govern an exemption from the
licensing requirements under proposed part 57.

[[Page 23643]]

This proposed requirement would be equivalent to that in Sec.
50.11(c). Proposed Sec. 57.45(c) would require issuance of a
construction permit, with the exception in proposed Sec. 57.45(d),
prior to starting construction of a utilization facility at a site and
would be equivalent to Sec. 50.10(c).
Proposed Sec. 57.45(d) would issue a general license for
construction activities on a site that is specified in a joint
application for a CP and associated OL(s) under proposed part 57 for a
nuclear reactor or nuclear plant subject to certain conditions in
proposed Sec. 57.45(d)(1)-(7). The proposed general license would
allow the general licensee to perform construction, as would be defined
in proposed Sec. 57.3, before NRC issuance of a construction permit
for the nuclear reactor or nuclear plant.
Proposed Sec. 57.45(d)(1) would require that the general licensee
has submitted, and the Commission docketed, a joint application for a
CP and associated OL(s) under proposed part 57. This proposed
requirement would include several additional conditions on the joint
application. First, the joint application would be required to
reference an ML issued by the Commission under 10 CFR chapter I. This
condition would provide assurance that the general licensee would not
complete construction of the nuclear reactor or nuclear plant before
issuance of the CP because the manufactured reactor would be an
essential part of the reactor or plant and proposed Sec. 57.45(d)(5)
would prohibit bringing it to the site under the general license.
Second, the joint application would be required to reference a CP and
OL issued pursuant to proposed part 57 that the Commission afforded
generic finality under proposed Sec. 57.142(e) and that referenced the
same ML as the general licensee's joint application. This condition
would ensure that the complete design had been reviewed and approved by
the NRC and that a nuclear reactor or nuclear plant of the same design
had been successfully constructed under NRC oversight and placed into
operation. This would also ensure that the public had been afforded an
opportunity for hearing on the design, including the postulated site
parameters for the design, in accordance with Sec. Sec. 57.142(e) and
57.60(c). Third, the joint application would be required to reference a
design that met the criteria for a categorical exclusion under proposed
subpart K of part 57. Taken together, the requirements proposed in
Sec. 57.45(d)(1)(i) and (ii) would provide assurance that the SSCs of
the nuclear reactor or nuclear plant, which could be difficult to
change after their construction, would not pose obstacles to eventual
issuance of an OL under proposed part 57. Fourth, proposed Sec.
57.45(d)(1)(iii) would require the joint application to include a plan
for redress of any adverse environmental impact from conduct of
activities under the general license should such redress be necessary.
This proposed requirement would be similar to the requirements in Sec.
50.10(d)(3)(iii), which requires a redress plan as part of an
application for a limited work authorization, and Sec. 50.11(b)(2),
which requires the Commission to consider redress of adverse
environmental impacts in determining whether to grant an exemption
permitting the conduct of construction activities prior to the issuance
of a construction permit.
Proposed Sec. 57.45(d)(2) would require that the general licensee
has notified the NRC under proposed Sec. 57.4 that all applicable
permits, licenses, approvals, and other entitlements in connection with
the proposed action that the general licensee was responsible for
obtaining have been obtained. Proposed Sec. 57.45(d)(3) would require
that applicable Federal environmental consultations have been
completed. This would ensure that construction activities would not
begin unless the NRC has the information it would need to fulfill its
obligations for environmental review under the AEA, NEPA, and other
relevant laws.
Proposed Sec. 57.45(d)(4) would require that the general licensee
not allow SNM or radioactive material that would be associated with the
operation of the nuclear reactor or nuclear plant under an operating
license issued pursuant to proposed part 57 to be brought to the site.
This would ensure that activities under the general license would not
create radiological hazards or irreversible radiological impacts at the
site that would otherwise be controlled by a CP or OL under proposed
part 57. This would also ensure that activities under the proposed
general license would not involve radiological security concerns. In
addition, proposed subpart P of part 26 would require implementation of
an appropriate FFD program during construction.
Proposed Sec. 57.45(d)(6) would require that the general licensee
allow for any NRC inspections that the Commission would deem necessary
related to activities that would be performed under the general
license. This would ensure that the NRC could apply experience gained
from inspection of the construction of the same nuclear reactor or
nuclear plant design if needed during construction activities that
would be conducted under the proposed general license.
Proposed Sec. 57.45(d)(7) would clarify that any activities
undertaken by the general licensee or on its behalf under the general
license would be entirely at the risk of the general licensee and would
have no bearing on the issuance of a construction permit under proposed
part 57 with respect to the requirements of the AEA, and rules,
regulations, or orders issued under the AEA. However, the general
licensee would be able to mitigate this additional regulatory risk
through careful site selection to ensure that site characteristics are
within the bounds of the postulated site parameters and by performing
construction activities following appropriate QA and FFD programs.
Based on the proposed requirements in Sec. 57.45(d)(1)-(7), the
Commission has determined that such general licensing would be for only
parts of utilization facilities, not constitute an unreasonable risk to
the common defense and security, and, therefore, be consistent with the
authority provided to the Commission by section 109(a) of the AEA.
Proposed Sec. 57.55, ``Content of applications; general
information,'' would provide general information requirements for the
content of joint applications under proposed part 57 and would be
equivalent to Sec. 50.33, ``Content of applications; general
information,'' with the exception that no emergency planning zones
would be defined for facilities licensed under proposed part 57.
Proposed Sec. 57.60, ``Contents of applications; technical
information,'' would provide technical information for the content of
joint applications and would be equivalent to Sec. 50.34, ``Contents
of applications; technical information,'' but would not include a
preliminary safety analysis report. Proposed Sec. 57.60(a) would
provide the technical requirements for an FSAR submitted as part of a
joint application under proposed part 57. Proposed Sec. 57.60(a)(1)(i)
would address the intended use of the reactor to include maximum power
and inventory of radioactive material. Proposed Sec. 57.60(a)(1)(ii)
would provide requirements for an FSAR to describe and assess safety
features and barriers designed into the facility to prevent or mitigate
the consequences of an accident similar to Sec. 50.34(a)(ii)(D)
without the requirement to comply with part 100 or the radiation dose
criterion for an individual in Sec. 50.34(a)(1)(ii)(D).
Proposed Sec. 57.60(a)(1)(iii) would require the applicant to
demonstrate, through an evaluation, that the dose-

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based entry criterion specified in proposed Sec. 57.25(a) is
satisfied.
Proposed Sec. 57.60(a)(1)(iv) through (vi) would require the
applicant to describe the design features associated with any remote or
autonomous operation or remote monitoring capabilities. Proposed Sec.
57.60(a)(1)(vii) would require the applicant to provide the analysis,
appropriate test programs, prototype testing, operating experience, or
a combination thereof that would demonstrate that each of the design
criteria attributes described by proposed Sec. 57.30 would be met.
Proposed Sec. 57.60(a)(2) would require the applicant to include
design basis and principal design criteria information in the
application including the relation of the design bases to the design
criteria, and the relation of the principal design criteria to the
design criteria attributes described in proposed Sec. 57.30. The
principal design criteria establish the necessary design, fabrication,
construction, testing, and performance requirements for safety-related
SSCs that provide reasonable assurance that the facility can be
operated without undue risk to the health and safety of the public. The
reference to principal design criteria in proposed Sec. 57.60(a)(2)
would not require the applicant to meet the General Design Criteria in
appendix A of part 50. However, the General Design Criteria in appendix
A could be generally applicable to other types of nuclear plants and
used as guidance in establishing the principal design criteria for a
facility using part 57.
This proposed rule would not impose QA requirements under existing
appendix B to part 50. Proposed Sec. 57.60(a)(3) would require the
applicant to describe its QA program to be applied to the design,
fabrication, manufacturing, construction, and testing of safety-related
SSCs and would be equivalent to Sec. 50.34(a)(7). Qualified suppliers
of nuclear-grade SSCs have decreased over the last several decades.
This shrinking base of suppliers, increasing demand for advanced
reactors, existing SSC upgrades and maintenance needs for the operating
fleet, restart of shutdown plants, and policies to buy U.S. products,
are creating a need for new suppliers to enter the market. At the same
time, the evolution of quality system requirements has led to the
development of several QA standards with shared elements. The NRC's
proposal to enable applicants to select QA programs could broaden the
supplier base and increase flexibility in procurement. This approach
may encourage participation from qualified commercial suppliers,
thereby expanding the pool of vendors available to support nuclear
projects. This could mitigate risks of shortages, backlogs, and higher
costs of deployment of microreactors and reactors with comparable risk
profiles.
Proposed Sec. 57.60(a)(4) would specify requirements related to
sites at which multiple nuclear reactors may be built or installed.
Proposed Sec. 57.60(a)(4)(i) and (ii) would require the applicant to
analyze and specify limits on the number and configuration of reactors
at the site and evaluate potential hazards to safety-related SSCs of
any operating reactors that could arise from activities associated with
construction, operation, and decommissioning of other reactors at the
site. These requirements would be similar to existing requirements in
Sec. 50.34(a)(11). Proposed Sec. 57.60(a)(4)(iii) would require the
joint application to include a description of the portions of the
nuclear plant that a nuclear reactor would share with one or more other
reactors over the lifetime of the plant and to specify the functional
requirements and measures to meet the requirements for any shared
safety-related SSCs. Proposed Sec. 57.60(a)(4)(iv) would require the
joint application to include technical specifications, as appropriate,
for shared portions of the nuclear plant.
Proposed Sec. 57.60(a)(5) would require the applicant to include
current and projected population distributions and site evaluation
factors for seismic, meteorological, hydrologic, and geologic
characteristics with appropriate consideration of natural phenomena.
The reason for establishing siting requirements would remain the same
as it has been historically, which is to ensure that licensees and
applicants assess what impact the site environs may have on a nuclear
plant (e.g., external hazards) and, conversely, what potential adverse
health and safety impacts a nuclear plant may have on nearby
populations in view of the site characteristics. Natural phenomena's
and site characteristics' impacts are key inputs into the design of
safety-related SSCs to ensure they can perform their intended safety
functions. The information required by proposed Sec. 57.60(a)(5) would
inform site selection demonstrating that the site characteristics would
be bounded by site parameters postulated for a given design.
Proposed Sec. 57.60(a)(6) would require the applicant to provide
an analysis and evaluation of safety-related SSCs related to
performance requirements and information that show that safety
functions will be accomplished and would be equivalent to Sec.
50.34(b)(2).
Proposed Sec. 57.60(a)(7) would require the applicant to provide
information on the kinds and quantities of radioactive materials
expected to be produced by operation and the means for controlling and
limiting radioactive effluents and radiation exposures within the
limits set forth in 10 CFR part 20 and would be equivalent to Sec.
50.34(b)(3). The application would have to include an estimate of the
quantity of each of the principal radionuclides expected to be released
annually to unrestricted areas in liquid effluents produced during
normal reactor operations, an estimate of the quantity of each of the
principal radionuclides of the gases, halides, and particulates
expected to be released annually to unrestricted areas in gaseous
effluents produced during normal reactor operations, and a description
of the equipment and procedures for the control of gaseous and liquid
effluents and for the maintenance and use of equipment installed in
radioactive waste systems.
Proposed Sec. 57.60(a)(8) would require the applicant to provide
information related to operational programs concerning facility
operations. These programs could be developed specifically for an
individual reactor or generically for a particular design to be
administered at a corporate or institutional level to support fleet
operations. Proposed Sec. 57.60(a)(8)(i)-(iii) would require the
applicant to include information related to the organizational
structure, training and qualification, conduct of operations, plans for
preoperational testing and initial operations, and plans for normal
operations, and would be equivalent to Sec. 50.34(b)(6)(i)-(iv).
Proposed Sec. 57.60(a)(8)(iv) would require emergency plans for
responding to an accidental release or loss of control of radioactive
material. Proposed Sec. 57.60(a)(8)(iv) would also require the
applicant to coordinate response needs with local emergency planning
and offsite response organizations. This proposed provision would
ensure adequate communication, coordination, and cooperation among
applicants, licensees, and offsite response organizations to establish
agreements and arrangements for offsite support and to ensure
protective measures can and will be taken as conditions warrant.
An emergency planning zone (EPZ) would not be defined for
facilities licensed under proposed part 57. An EPZ is most useful as a
planning tool for implementing precautionary actions through
predetermined, prompt protective measures to respond to

[[Page 23645]]

events that involve a wide-scale area involving multiple jurisdictions
and rapidly progressing incidents that could result in acute doses or
early health effects. The characteristics of facilities that would be
licensed under proposed part 57 provide assurance that planning for
such precautionary actions is unnecessary. Consistent with other NRC-
licensed facilities that do not have defined EPZs, the proposed rule
would ensure that applicants and licensees develop and maintain
capabilities to protect emergency workers and the public.
Proposed Sec. 57.60(a)(8)(v) would require the applicant to
describe its physical security program, cybersecurity program,
information security program, and access authorization program and is
equivalent to Sec. 50.34(c). The physical security program would need
to meet the security requirements in part 70. For radiological
sabotage, because these events could disrupt the performance of the
design of reactors licensed under proposed part 57, the applicant would
need to perform an assessment against the threat of radiological
sabotage. The purpose of this assessment would be to evaluate the
design against security events derived from the design basis threat
(DBT) of radiological sabotage defined in Sec. 73.1, ``Purpose and
scope,'' to determine if an operational program for physical security
is needed. The criterion for the assessment in proposed Sec.
57.60(a)(8)(v)(A)(3) would require an applicant to show that potential
consequences resulting from an event initiated by the DBT would result
in offsite doses below the values in Sec. 50.34(a)(1)(ii)(D) even if
mitigation and recovery actions, including any operator action, were
unavailable or ineffective. For those proposed part 57 applicants not
able to meet the criterion in proposed Sec. 57.60(a)(8)(v)(A)(3),
proposed subpart J would provide performance-based requirements for
licensees.
Proposed Sec. 57.60(a)(8)(v)(B) would require licensees to
establish, implement, and maintain a cybersecurity program in
accordance with either Sec. 73.54, ``Protection of digital computer
and communication systems and networks,'' or proposed Sec. 73.110,
``Cybersecurity program.'' Proposed Sec. 57.60(a)(8)(v)(C) would
require licensees to establish, implement, and maintain an information
protection system that complies with the requirements of Sec. Sec.
73.21, ``Protection of Safeguards Information: Performance
requirements,'' 73.22, ``Protection of Safeguards Information: Specific
requirements,'' and 73.23, ``Protection of Safeguards Information--
Modified Handling: Specific requirements,'' as applicable. Proposed
57.60(a)(8)(v)(D) would require licensees to establish, implement, and
maintain an access authorization program in accordance with Sec.
73.56, ``Personnel access authorization requirements for nuclear power
plants.''
Proposed Sec. 57.60(a)(8)(vi) would require the applicant to
provide proposed technical specifications prepared in accordance with
the requirements of Sec. 50.36, ``Technical specifications,'' and
would be equivalent to Sec. 50.34(b)(6)(vi).
Proposed Sec. 57.60(a)(8)(vii) would require the applicant to
submit procedures to be used to provide assurance that limiting
conditions for any operating reactors will not be exceeded as a result
of activities associated with the construction of any additional
reactors at the same site and would be equivalent to Sec.
50.34(b)(6)(vii).
Proposed Sec. 57.60(a)(8)(viii) would require the applicant to
provide a radiation protection program as part of its application and
would be similar to Sec. 20.1101, ``Radiation protection programs.''
Proposed Sec. 57.60(a)(8)(ix) would require the applicant to
provide a fire protection program and would be similar to Sec.
50.48(a). Proposed Sec. 57.60(a)(8)(ix)(A)-(C) would require the
applicant to describe the fire protection program for the facility, any
specific features necessary to implement the program, and an analysis
to demonstrate that a fire or explosion in any area of the plant would
not prevent a safety-related SSC from performing its safety function.
Proposed Sec. 57.60(a)(8)(ix)(D)-(H) would establish specific
requirements for the fire protection program.
Proposed Sec. 57.60(a)(8)(x) would require the applicant to
describe how the human factors engineering requirements of proposed
Sec. 57.395 would be addressed. Proposed Sec. 57.60(a)(8)(x) would
also require the applicant to describe the training, examination, and
proficiency programs necessary to meet the requirements of proposed
subpart P.
Proposed Sec. 57.60(a)(8)(xi) would require the applicant to
submit its description and plan for implementation of a remote
operation or monitoring program, if applicable. Remote operation and
remote monitoring are defined in proposed Sec. 57.3 as control of the
reactor and observation of plant data, respectively, from a location
outside of the site boundary. Stakeholders have expressed interest in
the incorporation of remote operation and monitoring into their plant
designs.
Proposed Sec. 57.60(a)(8)(xii) would require the applicant to
submit its program to ensure that systems and components meet the
requirements in the codes and standards identified in the application
in accordance with proposed Sec. 57.60(a)(9).
Proposed Sec. 57.60(a)(8)(xiii) would require the applicant to
submit its environmental qualification of safety-related electric
equipment and would be similar to Sec. 50.49(a), which requires an
applicant to establish a program for qualifying the electrical
equipment. ``Environmental qualification'' means the applicant would
assess possible degradation of safety-related SSCs by the effects of
various environmental conditions.
Proposed Sec. 57.60(a)(8)(xiv) would require the applicant to
describe its FFD program under part 26 and would be equivalent to Sec.
52.79(a)(44).
Proposed Sec. 57.60(a)(8)(xv) would require the applicant to
submit a staffing plan that details operations staffing and what
staffing will be available to provide other needed support functions as
proposed in Sec. 57.395(c).
Proposed Sec. 57.60(a)(8)(xvi) would allow the applicant to seek
approval of a plan for the storage of irradiated fuel after termination
of an OL and would be similar to Sec. 50.54(bb). The plan would need
to demonstrate compliance with all applicable irradiated fuel
possession, safety, and environmental requirements; include a plan for
funding the management of the fuel; and address, as applicable,
transportation of the irradiated fuel.
Proposed Sec. 57.60(a)(8)(xvii) would allow the applicant to seek
approval of a decommissioning plan by submitting its plan with its
joint application and would be similar to Sec. 50.82(b)(1), which
requires the submittal of a decommissioning plan to the Commission.
Proposed Sec. 57.60(a)(8)(xviii) would require the applicant to
describe the managerial and administrative controls to assure safe
operation. The managerial and administrative controls would promote
safe, reliable, and efficient plant operation, including related
maintenance activities. These controls would be in effect at all times
during the operational phase. These controls would be in the form of
procedures to effectively implement a QA program.
Proposed Sec. 57.60(a)(9) would require the applicant to provide
information on the use of codes and standards used to design the
facility. In proposed part 57, the NRC would not incorporate by
reference specific codes and standards

[[Page 23646]]

as is done under the existing regulations in Sec. 50.55a, ``Codes and
standards,'' because some codes and standards are technology specific.
Rather, the proposed rule would provide flexibility for the applicant
to choose which codes and standards, including generally recognized
consensus codes or standards to apply to the design of its facility.
The applicant would be required to name each proposed code or standard
and evaluate it for applicability, adequacy, and sufficiency.
Justification would need to be provided if the code or standard would
be supplemented or modified. Criteria from these consensus codes or
standards would need to be clearly stated and shown to provide the
appropriate level of reliability, safety, and performance capability.
The applicability of these criteria would need to be determined from
the safety assessment. However, the applicant could still choose to
utilize 10 CFR 50.55a. Proposed part 57 would allow for the use of
international codes and standards not previously used in NRC licensing,
but the NRC recognizes that the use of any consensus code or standard
would ultimately need to be found acceptable on an application-specific
basis during an individual licensing review.
Proposed Sec. 57.60(a)(10) would require the applicant to provide
analyses and descriptions of the equipment and systems for combustible
gas control required by paragraph (d) of Sec. 50.44, ``Combustible gas
control for nuclear power reactors,'' and would be similar to Sec.
50.34(g), ``Combustible gas control.''
Proposed Sec. 57.60(a)(11) would require applicants to demonstrate
their technical qualifications to carry out the proposed activities in
compliance with the regulations in 10 CFR chapter I. This requirement
would be similar to Sec. 50.34(a)(9).
Proposed Sec. 57.60(a)(12) would require applicants to provide a
description of the design-specific risk analysis methods used to
demonstrate adequate defense in depth and safety margins, along with
the results of that analysis. This approach would offer appropriate
flexibility for risk analysis methods to be developed and assessed
based on the application they are used to support. This would also
include consideration of how risk analysis results and insights are
relied upon, together with factors such as defense in depth, safety
margin, simplicity of design, and treatment of uncertainty.
Proposed Sec. 57.60(a)(13) would require an applicant to provide
information demonstrating how it will comply with requirements for
criticality accidents in Sec. 50.68, ``Criticality accident
requirements,'' with the exception that proposed Sec. 57.60(a)(13)
would limit the maximum nominal U-235 enrichment of fresh fuel
assemblies specified in Sec. 50.68(b)(7) to less than twenty (20.0)
weight percent to allow for the fuel enrichments anticipated for
reactors that would be licensed under proposed part 57.
Proposed Sec. 57.60(b) would require applicants to either justify
the use of a categorical exclusion or, if a categorical exclusion would
not apply, submit an environmental report, or an applicant-prepared
environmental assessment or environmental impact statement, in
accordance with 10 CFR part 51. Proposed Sec. 57.350(b) would
establish criteria under which certain NRC actions would be
categorically excluded from the requirement to prepare an environmental
assessment or environmental impact statement.
Proposed Sec. 57.60(c) would provide the option for an applicant
to include in its joint application a request for generic finality.
Under proposed Sec. 57.142(e) and Sec. 57.130(b)(7), affording the
licensee ``generic finality'' would mean that matters resolved in the
proceedings on the application for issuance of the CP and associated
OL(s) for which the applicant has requested and the Commission has
granted generic finality would be considered resolved in proceedings on
other joint applications that reference the approved CP or associated
OL(s). Proposed Sec. 57.60(c) would require the joint application to
include, in addition to the information that would be required by
proposed Sec. 57.60(a) and (b), site parameters postulated for the
design, including the design basis external hazard levels for the
relevant external hazards, and an analysis and evaluation of the design
in terms of those site parameters, and may include generic aspects of
operational programs and requirements of the types specified in
proposed Sec. 57.60(a)(8), to the extent practicable. This would
provide an alternate licensing pathway to an ML under proposed subpart
D for obtaining finality on a complete final design for a nuclear
reactor or nuclear plant. This would support high volume licensing of
designs of reactors that would be wholly constructed at the site of
operation and would also serve as a means for obtaining finality on the
design of the portions of a nuclear plant other than the manufactured
reactor, if one or more manufactured reactors were to be used.
Proposed Sec. 57.60(d) would provide the option for an applicant
to designate in its joint application for a CP and associated OL(s) a
large geographical area or areas, as opposed to a specific site or
sites, within which it proposes to construct and operate one or more
nuclear reactors. This proposed regulation would provide a licensing
pathway that could support rapid deployment of a reactor for disaster
relief or other time-critical application, or fleet deployment within a
large area. Proposed Sec. 57.60(d)(1)-(3) and (8) would require the
applicant to supplement the information under proposed Sec. 57.60(a)
and (b) to cover the entire designated area or areas, include maps, and
provide any restrictions on specific locations within the designated
area or areas.
Proposed Sec. 57.60(d)(4) would require a plan for storage of
irradiated fuel after termination of an operating license and proposed
Sec. 57.60(d)(5) would require the application to include a
decommissioning plan. Proposed Sec. 57.60(d)(6) would require the
application to include a procedure covering activities that will be
conducted in connection with constructing each reactor and placing it
into operation at a specific location. Together, these requirements
would ensure that the entire lifecycle of any nuclear reactor deployed
in this manner would be analyzed and subject to public hearing at the
construction permit review stage, thereby facilitating potential rapid
issuance of an operating license once a specific location is chosen and
the reactor constructed.
Proposed Sec. 57.60(d)(7) would require the application to include
a procedure that describes how the applicant would determine that a
specific location within a designated area is suitable for construction
and operation, including notification to the NRC, in the manner
specified under proposed Sec. 57.4, before beginning construction.
This procedure would provide assurance that any change in site
characteristics at a specific location within the designated area or
areas would be identified and verified to be within the bounds of the
site characteristics approved in the construction permit. The
notification that would be required by this procedure would allow the
NRC to conduct any inspections deemed necessary during construction and
prepare for activities needed to make the finding required by proposed
Sec. 57.100(b)(1) and issue an OL.
Proposed Sec. 57.80, ``Standards for review of applications,''
would require a joint application for a CP and associated OL(s) to be
reviewed under the standards in parts 20, 50, 51, 54, 55, 70, 71, 72,
73, 74, and 140, as applicable, and that the Commission must perform an
environmental review of the application in accordance with

[[Page 23647]]

the provisions in proposed subpart K of part 57 and part 51.
Paragraphs (a) through (i) of proposed Sec. 57.90, ``Common
standards for licenses,'' would establish requirements for standards
that the NRC would consider in determining whether a CP or OL under
part 57 would be issued to an applicant. These requirements would be
equivalent to those in Sec. Sec. 50.23, ``Construction permits,''
50.40, Common standards,'' 50.42, ``Additional standard for class 103
licenses,'' 50.43(a)-(d), 50.45, ``Standards for construction permits,
operating licenses, and combined licenses,'' and 50.50, ``Issuance of
licenses and construction permits,'' except proposed Sec. 57.90(h)
would specify that a CP would be converted into one or more OLs.
Proposed Sec. 57.95, ``Issuance of construction permit,'' would
address issuance of construction permits, such as the findings the
Commission must make, the authorization provided by the construction
permit, and limits on that authorization. Proposed Sec. 57.95(a) is
based on Sec. 52.97, ``Issuance of combined licenses,'' which covers
issuance of combined licenses because under proposed part 57, the
Commission would review the final design and any operational programs
and requirements that are material to the adequacy of the design as
part of the construction permit review. Unlike Sec. 52.97(a)(1)(iii),
proposed Sec. 57.95(a)(3) would not include a finding about whether
the facility would operate in conformity with the license as this would
be left for the issuance of the OL under proposed Sec. 57.100,
``Issuance of operating license.'' Proposed Sec. 57.95(b) would be
equivalent to Sec. 50.35(b), except that it would specify that the
construction permit would not constitute Commission approval of the
operational programs and requirements provided in the application
unless the applicant specifically requests such approval and such
approval is incorporated in the construction permit. Proposed Sec.
57.95(c) would be equivalent to Sec. 50.35(c).
Proposed Sec. 57.100, ``Issuance of operating license,'' would
address issuance of OLs, such as the findings the Commission must make,
requests for low power testing, and conditions on the OL. Proposed
Sec. 57.100(a) would be equivalent to Sec. 50.56, ``Conversion of
construction permit to license; or amendment of license.'' Proposed
Sec. 57.100(b)(1) through (6) would be equivalent to Sec. 50.57(a)(1)
through (6). Proposed Sec. 57.100(c) would be equivalent to 50.57(b).
Proposed Sec. 57.100(d) would be equivalent to 50.57(c).
Proposed Sec. 57.100(e) would require an operating license that
references an ML to include a condition, as appropriate, that would
specify that the authorization to operate the reactor would be
suspended while features to prevent criticality are in place. The
condition would also specify that initiation of removal of features to
prevent criticality would not be allowed unless either all conditions
of an OL issued under proposed part 57 authorizing operating of the
reactor were satisfied, or the reactor had been defueled in accordance
with an appropriate license issued by the Commission.
Proposed Sec. 57.100(f) would specify that an OL for a nuclear
reactor that would be part of a nuclear plant at which portions of the
nuclear plant would be shared with one or more other reactors over the
lifetime of the plant as described in proposed Sec. 57.60(a)(4)(iii),
must include a condition specifying that the shared portions of the
plant would be part of the facility as described in the operating
license's FSAR and any related technical specifications under proposed
Sec. 57.60(a)(4)(iv) would be incorporated in the license. This
proposed requirement would ensure that shared portions of a nuclear
plant and any shared safety-related SSCs would be appropriately
considered in each OL for a nuclear reactor that would be part of the
nuclear plant and support the requirements in proposed Sec. 57.305,
``Decommissioning and license termination,'' for decommissioning a
nuclear plant at which more than one reactor would be operated over the
lifetime of the plant.
Proposed Sec. 57.105(a) would address the duration of a CP and OL
and would be equivalent to Sec. 50.51(a). Proposed Sec. 57.105(b)
would address cessation of operations and the continued possession and
ownership of the nuclear reactor or nuclear plant and would be
equivalent to Sec. 50.51(b).
Proposed Sec. 57.110, ``Transfer of licenses,'' would establish
requirements for the transfer of a CP or OL by providing the equivalent
requirements of Sec. 50.80, ``Transfer of licenses.''
Proposed Sec. 57.115, ``Application for renewal,'' would address
applications for renewal of OLs. Proposed Sec. 57.110(a) would require
the filing of an application for a renewed license to be in accordance
with proposed Sec. Sec. 57.4 and 57.7. Proposed Sec. 57.115(b)-(e)
would specify the information required to be included in an application
for renewal to include the technical specifications and information
related to general, technical, environmental, and aging management
requirements and would be equivalent to Sec. Sec. 54.19, ``Contents of
application--general information,'' 54.21, ``Contents of application--
technical information,'' and 54.22, ``Contents of application--
technical specifications,'' albeit modified to reflect the requirements
for the FSAR, environmental report, and technical specifications for
reactors licensed under proposed part 57. Proposed Sec. 57.115(f)
would address hearing opportunities and would be equivalent to Sec.
54.27, ``Hearings.''
Proposed Sec. 57.120, ``Criteria for renewal,'' would address the
Commission's criteria for issuing a renewed operating license and would
be equivalent to Sec. 54.29, ``Standards for issuance of a renewed
license.''
Proposed Sec. 57.130, ``Hearings,'' would address requirements for
hearings for CPs and OLs and would be equivalent to the requirements in
Sec. 50.58(b) and Sec. 54.27. If an applicant were to request generic
finality under proposed Sec. 57.60(c), then the Commission's ruling on
a request for hearing or petition for leave to intervene under 10 CFR
2.309(d)(2) would consider that a petitioner may have an interest that
may be affected by the proceeding on the application if matters
resolved in the licensing proceeding were to be afforded generic
finality under proposed Sec. 57.142, ``Finality for construction
permits and operating licenses.'' This would enable petitioners whose
property, financial, or other interests would not be directly affected
by the issuance of the CP and OL for a particular reactor to have an
opportunity to intervene on generic aspects of the design that would be
afforded finality and would therefore not be subject to hearing if
referenced in a joint application for a CP and associated OL(s) that
would affect the petitioner's property, financial, or other interest.
Proposed Sec. 57.130(b)(7) would require the Commission to include an
applicant's request for generic finality as a proposed action in the
joint notice of hearing and proposed action that would be required by
Sec. Sec. 2.104, ``Notice of hearing,'' and 2.105, ``Notice of
proposed action.''
Proposed Sec. 57.135, ``Duration of renewal,'' would require that
a renewed OL be issued for a fixed period of time beyond the expiration
of the current OL. The period would be the sum of the amount of time
beyond the expiration of the OL requested in a renewal application plus
any remaining years on the operating license currently active. This
proposed rule would provide that no renewed license would exceed more
than 40 years in duration, which is limited by the AEA.

[[Page 23648]]

Proposed Sec. 57.142 would include requirements to address
finality for construction permits and operating licenses and would be
similar to the finality provisions for MLs in proposed Sec. 57.175,
``Finality of manufacturing licenses; information requests.'' Proposed
Sec. 57.142(e) would specify that the Commission may afford generic
finality to generic aspects of the design of a nuclear reactor or
nuclear plant, including postulated site parameters, and generic
operational programs and requirements submitted pursuant to proposed
Sec. 57.60(c), if it finds that the proposed generic design can be
constructed and operated at sites having characteristics that fall
within the site parameters postulated for the design, and in accordance
with the generic operational programs and requirements, without undue
risk to the health and safety of the public. This proposed requirement
would provide an alternative to an ML for standardization of nuclear
reactor or nuclear plant designs and operational programs and
requirements for the purpose of referencing in a subsequent joint
application for a CP and associated OL(s) under proposed part 57.

E. Subpart D--Manufacturing Licenses

Provisions related to MLs were first adopted by the NRC in 1973
through the addition of appendix M to part 50. The regulation supported
the manufacture of a nuclear power reactor to be incorporated into a
commercial nuclear plant under a CP and operated under an OL at a
different location from the place of manufacture. The regulations and
processes for MLs were changed substantially in the part 52 rulemaking
in 2007 (72 FR 49352). The most important shift in the ML concept in
that rulemaking was that a final reactor design, which would be
equivalent to that required for a standard design certification under
part 52 or an OL under part 50, must be submitted and approved before
issuance of an ML. The rationale for that change was that approval of a
final design ensures early consideration and resolution of technical
matters before there is any substantial commitment of resources
associated with the actual manufacture of the reactor, which greatly
enhances regulatory stability and predictability.
Proposed subpart D would address applications for, issuance of, and
other provisions related to MLs covering manufacturing activities at
one or more licensee facilities under proposed part 57. These proposed
requirements would be largely equivalent to those in part 52 for MLs.
Proposed Sec. 57.145, ``Scope,'' would address the scope of the
proposed subpart D sections and would be equivalent to Sec. 52.151,
``Scope of subpart,'' except that it also would state that the scope of
proposed subpart D includes requirements for manufacturing manufactured
reactors at a manufacturing facility, loading fuel into manufactured
reactors at the manufacturing facility, and transportation of
manufactured reactors.
Proposed Sec. 57.150, ``Contents of applications for manufacturing
licenses; general information,'' would address general information
requirements for the content of ML applications and would be equivalent
to Sec. 52.156, ``Contents of applications; general information,''
with one exception. Proposed Sec. 57.150 would require each
application for an ML to also include the information required by
proposed Sec. 57.55(e). This information would include the type of
license applied for, the use to which the facility will be put, the
period of time for which the license is sought, and a list of other
licenses, except operator's licenses, issued or applied for in
connection with the proposed facility to address the potential
variations in how MLs might be formulated under proposed part 57.
Proposed Sec. Sec. 57.155, ``Contents of applications; technical
information in final safety analysis report,'' and 57.160, ``Contents
of applications; additional information,'' would address requirements
for the technical content of applications for MLs to be included in the
FSAR and additional information to be included in the application and
would be equivalent to Sec. Sec. 52.157, ``Contents of applications;
technical information in final safety analysis report,'' and 52.158,
``Contents of applications; additional technical information,'' with
three significant exceptions. First, proposed Sec. 57.155(c) would
include the option for the application to include final, non-site-
specific design information for a nuclear plant that would use a
reactor manufactured under the ML. This would allow the NRC to review
the design of the entire nuclear plant and afford finality in
accordance with proposed Sec. 57.175, which would increase the
efficiency of reviewing a joint application for a CP and associated
OL(s) under proposed subpart C that references the ML. Second, proposed
Sec. 57.155 would not include a requirement for proposed inspections,
tests, analyses, and acceptance criteria to be included in the
application because they would not be required for the issuance of OLs
under proposed subpart C. Third, proposed Sec. 57.160(a) would provide
the option for an applicant to include in its application descriptions
of generic operational programs and requirements, which the NRC could
afford finality to in accordance with proposed Sec. 57.175.
In addition, the requirements in proposed Sec. Sec. 57.155 and
57.160 would be modified from the analogous requirements in Sec. Sec.
52.157 and 52.158 to align with the technical requirements in proposed
part 57. Proposed Sec. 57.155(a) would outline the required content of
the application addressing design information and state that the
application must include design information equivalent to that required
for a joint application for a CP and associated OL(s) under proposed
subpart C, other than site-specific information, relevant to the
manufactured reactor.
Proposed Sec. 57.160(b) would require an ML application to include
either the information justifying application of a categorical
exclusion as described in proposed subpart K of part 57, or an
environmental report or applicant-prepared environmental assessment, in
accordance with 10 CFR part 51.
Proposed Sec. 57.160(c) would require an ML application to include
a description of the safeguards information program, in accordance with
Sec. Sec. 73.21 and 73.22 of this chapter, as applicable, to prevent
any unauthorized disclosure.
Proposed Sec. 57.160(d)(1) would require an ML application to
include a description of the relevant codes and standards used in the
procurement, fabrication, and assembly of components comprising the
manufactured reactor. Proposed Sec. 57.160(d)(2) would require an ML
application to include a description of the organizational and
management structure responsible for the design and manufacturing of
the manufactured reactor. Proposed Sec. 57.160(d)(3) would require an
ML application to include a description of the tests and inspections to
be performed during the manufacturing and fabrication process,
including components, as well as an assembled manufactured reactor.
Proposed Sec. 57.160(d)(4) would require an ML application to include
a description of the fitness-for-duty program required by part 26.
Proposed Sec. 57.160(e) would provide application requirements
related to the deployment of the completed manufactured reactor.
Proposed Sec. 57.160(e)(1) would require inclusion of information
related to the procedures governing the preparation of the manufactured
reactor for shipping to the site where it is to be operated, the
conduct of shipping, and the verification of the condition of the

[[Page 23649]]

shipped items upon receipt at the site. Proposed Sec. 57.160(e)(2)
would require that the application include information on the
interaction of the design, manufacture, and installation of a
manufactured reactor within the applicant's organization and the manner
by which the applicant would ensure close integration between the
designer, contractors, and any licensee of a facility in which the
manufactured reactor is to be installed. Finally, proposed Sec.
57.160(e)(3) would require that the application include a description
of the measures to be used for the control of interfaces between the
holder of the ML and the holder of the CP for the nuclear plant at
which the manufactured reactor is to be installed. This information
would be necessary for the NRC to determine whether the applicant has
appropriate controls in place to ensure coordination between parties
involved in the design, manufacture, and eventual operation of any
reactor manufactured under an ML.
Proposed Sec. 57.160(f) would include additional requirements for
application content for applicants seeking an ML for manufactured
reactors that will be fueled at the manufacturing facility under a
license issued in accordance with 10 CFR part 70, ``Domestic Licensing
of Special Nuclear Material,'' consistent with the requirements in
proposed Sec. 57.197(d). These provisions would require the
application to include information related to loading fuel and the
required features to prevent criticality and to otherwise provide
assurance that the fueled manufactured reactor could be successfully
transported, installed, and operated at a site for which the Commission
has issued a CP under proposed subpart C that authorizes construction
of a nuclear plant using the manufactured reactor.
Proposed Sec. Sec. 57.165, ``Standards for review of
applications,'' and 57.170, ``Administrative review of applications;
hearings,'' would provide standards for review of applications and
administrative review of applications for MLs, including hearings, and
would be equivalent to Sec. Sec. 52.159, ``Standards for review of
applications,'' and 52.163, ``Administrative review of applications;
hearings.''
Proposed Sec. 57.172, ``Issuance of manufacturing license,'' would
address issuance of an ML and would be equivalent to Sec. 52.167,
``Issuance of manufacturing license,'' with two exceptions. First,
proposed Sec. 57.172(a)(6) would include a requirement that the
Commission make a finding that generic operational programs submitted
as part of the ML application under proposed Sec. 57.160(a) provide
reasonable assurance that the manufactured reactor can be operated
under an operating license that references the manufacturing license in
conformity with the provisions of the AEA and the Commission's
regulations. Second, proposed Sec. 57.172(b)(4) would require each ML
issued under proposed part 57 to specify that the portions of the
nuclear plant other than the manufactured reactor must be as described
in the information included in the ML application if the applicant
chose to include this information in accordance with proposed Sec.
57.155(c)(8) instead of interface requirements. These provisions of
proposed Sec. 57.172 could greatly reduce the scope of and timeframe
for review of a joint application for a CP and associated OL(s) that
references the ML because the NRC would have afforded finality to the
entire nuclear plant design and potentially nearly all the operational
programs through the ML proceeding, allowing the review of the joint
application to focus on site-specific information.
Proposed Sec. 57.175 would address finality of MLs and would be
equivalent to Sec. 52.171, with the exception that proposed Sec.
57.175(d) would allow the holder of an ML to use the regulations in
Sec. 50.59, ``Changes, tests, and experiments,'' to determine whether
changes to the facility or procedures as described in the FSAR would
require an amendment to the ML. This would be different than the
provisions in Sec. 52.171 that do not allow any changes to the design
of a manufactured reactor without requesting a license amendment.
Proposed Sec. 57.180, ``Duration of manufacturing license,'' would
address the duration of MLs. However, compared to the current analogous
requirements in Sec. 52.173, ``Duration of manufacturing license,''
proposed Sec. 57.180 would not include a minimum duration for an ML
and would provide for a 40-year maximum for the duration of an ML.
These differences would be consistent with the requirement in proposed
Sec. 57.55(e) that each application must state the period of time for
which the license is sought and the limitation on the duration of
design certifications in Sec. 52.55, ``Duration of certification.''
Proposed Sec. 57.185, ``Transfer of manufacturing license,'' would
address the transfer of MLs and would be equivalent to Sec. 52.175,
``Transfer of manufacturing license.''
Proposed Sec. 57.190, ``Renewal of manufacturing licenses,'' would
address the renewal of MLs and would be equivalent to Sec. Sec.
52.177, ``Application for renewal,'' 52.179, ``Criteria for renewal,''
and 52.181, ``Duration of renewal,'' with a minor exception. Proposed
Sec. 57.190(b) would state that an ML for which a timely application
for renewal has been filed would remain in effect until the Commission
has made a final determination on the renewal application. However,
this provision would omit a limitation from the equivalent provision in
Sec. 52.177, which prohibits the holder of an ML from beginning the
manufacture of a manufacture reactor less than 3 years before the
expiration of the license. This limitation would be omitted because
applicants under proposed part 57 may present smaller, simpler designs
in ML applications than those that were envisioned when the existing
requirements were written. Eliminating the 3-year constraint in this
provision would provide greater flexibility for ML holders related to
manufactured reactors being produced close to the time when the ML
expires. Finally, proposed Sec. 57.190(e) would provide for a 40-year
term for a renewed ML, consistent with the term for an initial ML under
proposed Sec. 57.180.
Proposed Sec. 57.197, ``Manufacturing,'' would include
requirements covering the activities performed under an ML issued under
proposed part 57. Proposed Sec. 57.197 would also include requirements
that apply to portions of a manufactured reactor in recognition that
some activities covered by an ML may occur at different fabrication
facilities. Proposed Sec. 57.197(a) would establish the requirements
to have in place programs, procedures, and a well-defined command and
control structure to manage manufacturing-related activities.
Proposed Sec. 57.197(b) would include requirements for executing
the manufacturing activities following receipt of an ML under proposed
part 57. These requirements would include conducting manufacturing
processes within facilities for which the license holder can control
access and activities that might affect manufacturing, performing
manufacturing in accordance with the ML and appropriate codes and
standards, and establishing and implementing post-manufacturing
inspections.
Proposed Sec. 57.197(c) would provide requirements for the control
of radioactive materials if the holder of an ML plans to possess and
use source, byproduct, or special nuclear material as part of the
manufacturing process. By and large, the proposed Sec. 57.197 would
refer to NRC regulations in 10 CFR part 30, ``Rules of General
Applicability to Domestic Licensing of Byproduct Material,'' 10 CFR
part 40, ``Domestic

[[Page 23650]]

Licensing of Source Material,'' and part 70 for the requirements on
controlling radioactive materials. The NRC proposes several specific
requirements to address the potential hazards of radioactive materials
in areas such as having a fire protection program, an emergency plan,
training programs, and procedures to minimize contamination.
The most significant change proposed for MLs in part 57 (which
would be similar to changes for MLs under part 53) as compared to MLs
under part 52 relates to proposed Sec. 57.197(d), which would allow
and establish requirements for the loading of fuel into a manufactured
reactor at the manufacturing site for subsequent transport to a nuclear
plant that would be constructed pursuant to a CP that would be issued
under proposed part 57. The first requirement in proposed Sec.
57.197(d) would establish limitations on when a holder of an ML under
proposed part 57 and a license under part 70 could load fuel into a
reactor manufactured under the ML. The proposed regulation would
require that features to prevent criticality specified in the ML be in
place before loading fuel into the manufactured reactor and during the
reactor's storage and transport. The proposed requirement would provide
flexibility because of the potential variety of reactor designs, the
variety of possible measures to prevent criticality, and the range of
possible conditions associated with the loading of fuel into, storage
of, and transport of manufactured reactors. For example, the features
to prevent criticality that could be considered individually and
collectively to address possible adverse conditions include the
reactivity control systems in place to support operations, inherent
features of the fuel and materials within a manufactured reactor, and
temporary measures or physical mechanisms (e.g., neutron poisons) for
specific circumstances and conditions. This proposed requirement would
contribute to the NRC's longstanding practice of requiring defense in
depth for preventing accidents in any facility possessing or using SNM,
including requirements in Sec. 70.22(a)(8) for procedures to protect
health and minimize danger to life or property (e.g., procedures to
avoid accidental criticality, determine subcritical limits on
controlled parameters under normal conditions or subcritical values
under abnormal conditions, monitor personnel and waste disposal,
provide post-criticality accident emergency response, and adhere to the
double contingency principle where practicable).
The proposed requirements to have in place features to prevent
criticality could likewise support meeting other provisions in part 70,
such as those related to equipment and procedures that protect health
and minimize danger to life or property. The features to prevent
criticality in the proposed part 57 requirements would reasonably
ensure that a manufactured reactor does not become critical over a
range of possible conditions. With the requirements for features to
prevent criticality under proposed part 57 and all criticality safety
controls required by part 70 in place, the presence of fuel in the
manufactured reactor would not create a nuclear hazard different than
the hazard from the presence of the same fuel in a storage location or
container licensed under part 70. Collectively, these measures would
reasonably ensure that the manufactured reactor is not capable of
operations, thereby obviating the need for an OL under proposed subpart
C of part 57 to authorize fuel loading. Additionally, this approach
would focus the ML application and its review on the design,
manufacture, and deployment of the manufactured reactor.
The activities involving SNM within the manufacturing facility,
including the loading of fuel, would be regulated primarily under the
part 70 license. The provisions of subpart H to part 70 would not be
applicable to a part 70 license that only authorizes possession of
special nuclear material for the purpose of loading fresh fuel into a
manufactured reactor. The reference to the requirements in part 70 in
proposed Sec. 57.197(d) would reasonably assure that the applicant
will utilize the appropriate equipment and procedures to protect health
and minimize danger to life or property. The regulations in part 51
provide a flexible approach for environmental review to address the
range of regulated activities under part 70. The flexibility in part 51
would enable the NRC to determine the appropriate type of environmental
review based on the circumstances associated with the loading of fuel
into a specific manufactured reactor.
Proposed Sec. 57.197(d) would cite the requirements in 10 CFR
parts 70 and 73 to ensure important features and programs are in place
prior to the receipt of SNM. The features and programs that would be
required by 10 CFR parts 70 and 73 to be in place prior to receipt of
SNM would include (1) radiation monitoring instrumentation and alarms;
(2) measures to detect potential criticality accidents; (3) appropriate
procedures, equipment, and personnel qualified for the fuel loading;
(4) programs for physical security and cybersecurity; and (5) material
control and accounting (MC&A) programs.
Proposed Sec. 57.197(d)(2) would cover the activities related to
the storage, movement, and loading of fresh fuel into a manufactured
reactor in the manufacturing facility and would likewise refer to the
applicable regulations in part 70.
Proposed Sec. 57.197(d)(3) would include requirements to address
security programs for any ML authorizing possession of a manufactured
reactor into which fuel has been loaded at the manufacturing facility.
Currently, for category II SNM, security measures may be required in
addition to requirements included in Sec. 73.67, ``Licensee fixed site
and in-transit requirements for the physical protection of special
nuclear material of moderate and low strategic significance,'' on a
case-by-case basis. Including appropriate security measures in the
proposed part 57 regulations would provide additional openness and
transparency for applicants applying for an ML who seek to load fuel
into manufactured reactors at a manufacturing site.
Currently, Sec. 73.67 only requires a security plan for licensees
who possess, use, transport, or deliver to a carrier for transport SNM
of moderate strategic significance, or 10 kg or more of SNM of low
strategic significance. However, the physical security program for
fueled manufactured reactors would require a security plan for any ML
authorizing possession of a manufactured reactor into which fuel has
been loaded at the manufacturing facility, regardless of fuel type,
enrichment, and quantity. This would be consistent with other controls
proposed for MLs, including reactivity and criticality controls.
The proposed Sec. 57.197(d)(3) would also require a holder of an
ML that would load fuel into a manufactured reactor under a part 70
license to address cybersecurity to ensure a cyberattack would not
adversely impact the functions performed by digital assets necessary
for physical security, radiation monitoring, or criticality prevention.
Proposed Sec. 57.197(d)(4) would require the loading or unloading
of fuel into or from a manufactured reactor and any changes to the
configuration of reactivity-related systems to be performed by a
certified fuel handler.
Proposed Sec. 57.197(e) would only allow the transport or removal
of a manufactured reactor or portions of a manufactured reactor for
either (1) delivery to a domestic site for which the

[[Page 23651]]

Commission has issued a CP authorizing the construction of a nuclear
plant using a manufactured reactor under the specific ML, or (2) export
in accordance with 10 CFR part 110, ``Export and Import of Nuclear
Equipment and Material.'' This proposed requirement would be similar to
the limitations in Sec. 52.153, ``Relationship to other subparts,''
with the difference being that proposed part 57 would allow the
installation of a manufactured reactor only at the site of a CP issued
under proposed subpart C of part 57. An additional paragraph in
proposed Sec. 57.197(e) would provide requirements for protecting
fueled manufactured reactors during transport to the site of the
nuclear plant by referencing the transportation and security
requirements in 10 CFR part 71 and part 73. As previously noted,
proposed Sec. 57.197(e) would include an additional provision that
would allow a manufactured reactor or portions of a manufactured
reactor to be removed from the place of manufacture for export in
accordance with 10 CFR part 110, which represents another difference
from the similar provision in Sec. 52.153.
Proposed Sec. 57.197(f) would include requirements for the
acceptance of a manufactured reactor at the site of a nuclear plant
specified in a CP issued under proposed subpart C of part 57 and would
require that the manufactured reactor be installed in accordance with
that CP. Other requirements in proposed Sec. 57.197(f) would address
required receipt inspections and verification that any interface
requirements between the manufactured reactor and the balance of the
nuclear plant have been met.

F. Subpart E--Standard Design Approvals

Proposed subpart E would address applications for, issuance of, and
other requirements related to SDAs under proposed part 57. Proposed
Sec. 57.200, ``Scope,'' would describe how the contents of proposed
subpart E would address SDAs and would be equivalent to Sec. 52.131,
``Scope of subpart.'' Proposed Sec. 57.205, ``Contents of
applications; general information,'' would address general information
requirements for the content of applications and would be equivalent to
Sec. 52.136, ``Contents of applications; general information.''
Proposed Sec. 57.210, ``Contents of applications; technical
information,'' would address requirements for the technical content of
applications and would be largely equivalent to Sec. 52.137,
``Contents of applications; technical information.'' Proposed Sec.
57.210 would include additional requirements for applications for
approval of a ``major portion'' of a standard design. Additional
discussion regarding standard design approvals for a major portion of a
standard design can be found in the NRC's ``A Regulatory Review Roadmap
for Non-Light Water Reactors,'' which considers the Nuclear Innovation
Alliance report, ``Clarifying `Major Portions' of a Reactor Design in
Support of a Standard Design Approval.'' Proposed Sec. 57.210(a) would
outline the required content of the FSAR. This content would be
modified from the analogous requirements in Sec. 52.137 to align with
the technical requirements in proposed part 57. Proposed Sec.
57.210(b)(1) for portions of the application addressing design
information would state that the application must include design
information equivalent to that required for a joint application for a
CP and associated OL(s) under proposed subpart C, other than site-
specific information, relevant to the scope of the SDA.
Proposed Sec. 57.213, ``Standards for review of applications,''
would address standards for review of applications and would be
equivalent to Sec. 52.139, ``Standards for review of applications.''
Proposed Sec. Sec. 57.215, ``Staff approval of design,'' would address
staff approval of designs and would be equivalent to Sec. Sec. 52.143,
``Staff approval of design.''
Proposed Sec. 57.220, ``Finality of standard design approvals;
information requests,'' would address finality of standard design
approvals and information requests and would be equivalent to Sec.
52.145, ``Finality of standard design approvals; information
requests.'' There would be no equivalent to proposed Sec. 57.220(d) in
part 52 for standard design approvals. This provision would state that
the Commission will require, before granting a CP, OL, or ML that
references a standard design approval, that information normally
contained in engineering documents be completed and available for
audit. A similar provision is included in Sec. 52.47, ``Contents of
applications; technical information,'' in relation to a standard design
certification. Proposed Sec. 57.220(d) would require that design and
analysis information that would be needed for the Commission to make
its safety determination be complete and available for any application
the NRC would be reviewing. Making this explicit would provide
increased clarity to future standard design approval applicants under
proposed part 57.
Proposed Sec. 57.225, ``Duration of design approval,'' would
specify that an SDA under the part 57 framework does not expire, which
is different than the current regulation in Sec. 52.147, ``Duration of
design approval,'' that limits the validity of an SDA under the part 52
framework to 15 years and prohibits renewal. Proposed Sec. 57.220(a)
would specify that the NRC staff and the ACRS do not have to use or
rely on the earlier determination on an SDA under the proposed Sec.
57.215 in their review of any application under proposed part 57 that
incorporates by reference the SDA if there exists significant new
information or for other good cause that substantially affects the
earlier determination. This would allow the NRC staff and ACRS to
address potential issues, including but not limited to design
obsolescence or advances in the state of the art, that might arise
because of the indefinite duration of the SDA. This change would also
reduce the administrative burden on applicants and the NRC associated
with a request for re-approval of a standard design and would align
with the indefinite validity (as supported by renewals) of OLs and MLs
that could reference an SDA.

G. Subpart F--Reporting of Defects and Noncompliance

Proposed subpart F of part 57 would establish procedures and
requirements for implementation of section 206 of the Energy
Reorganization Act of 1974. That section requires any individual
director or responsible officer of a firm constructing, owning,
operating, or supplying the components of any facility or activity that
is licensed or otherwise regulated pursuant to the AEA or the Energy
Reorganization Act of 1974, to immediately notify the Commission if
they obtain information reasonably indicating certain failures to
comply or defects, unless the individual has actual knowledge that the
Commission has been adequately informed of the failure to comply or
defect. These failures to comply or defects are the following: the
facility, activity, or basic component supplied to such facility or
activity fails to comply with the AEA or any applicable rule,
regulation, order, or license of the Commission relating to substantial
safety hazards; or the facility, activity, or basic component supplied
to such facility or activity contains defects that could create a
substantial safety hazard.
The proposed Sec. 57.240, ``Definitions,'' would provide
definitions that are consistent with those applicable to non-power
reactors in 10 CFR part 21, ``Reporting of Defects and Noncompliance,''
with some slight differences to be technology neutral and reflect the
types of facilities that would be eligible for licensing under proposed

[[Page 23652]]

part 57. The proposed definition of ``Basic component'' would be
slightly different than the definition in Sec. 50.2 in that the
proposed definition would cover the same concept but would be
technology neutral and reference the accident dose entry criterion in
proposed Sec. 57.25(a). The proposed Sec. 57.240 would specifically
define ``construction'' or ``constructing'' for use in proposed subpart
F to mean the analysis, design, manufacture, fabrication, placement,
erection, installation, modification, inspection, or testing of a
facility or activity that is subject to the regulations in proposed
part 57 and safety-related consulting services related to the facility
or activity. This definition of ``constructing'' or ``construction''
would be different than the definition in proposed Sec. 57.3 because
it is needed to define the applicability of proposed Sec. 57.240 and
part 21. The proposed definition of ``Dedicating entity'' is slightly
different than the definition in Sec. 21.3. The proposed definition
would state that the dedicating entity would be the organization that
performs the dedication process and would not otherwise describe the
dedicating entity like in Sec. 21.3. The proposed definition of
``Dedication'' is slightly different than the definition in Sec. 21.3.
The dedication process must be conducted in accordance with the
applicant's applicable provisions for their proposed Sec. 57.60(a)(3)-
required quality assurance program rather than appendix B to part 50.
Proposed Sec. 57.270, ``Notification of failure to comply or
existence of a defect and its evaluation,'' would require the holders
of construction permits and manufacturing licenses under proposed part
57 to report any significant breakdown in quality assurance and would
be equivalent to requirements in Sec. 50.55(e). Proposed Sec. 57.285,
``Maintenance and inspection of records,'' would provide record
retention requirements for the holders of construction permits and
manufacturing licenses under proposed part 57 that would be equivalent
to record retention requirements in Sec. 50.55(e). All other sections
of proposed subpart F would be equivalent to corresponding part 21
provisions.

H. Subpart G--Irradiated Fuel Storage, Decommissioning, and License
Termination Requirements

1. Irradiated Fuel Storage
The NRC proposes to regulate irradiated fuel storage by entities
licensed under proposed part 57 by requiring a combination of a license
under 10 CFR part 70, a general or site-specific license under 10 CFR
part 72, and the use of a certified irradiated nuclear fuel dry storage
system under part 72.
The NRC proposes to issue to the holder of an OL under proposed
part 57 a part 72 general license for the disposition of irradiated
fuel, similar to the general license issued to the holder of a part 50
OL under Sec. 72.210, ``General license issued.'' Proposed Sec.
57.300(a) would permit the proposed part 57 OL holder to store the
irradiated fuel from its reactor at the operating site within the
reactor or in an irradiated fuel storage system certified under part
72. The NRC proposes to allow in-reactor storage of irradiated fuel
because the conditions of the reactor are essentially unchanged whether
the reactor is in operation or has ceased operations (e.g., radiation
shielding, confinement, passive heat dissipation). Thus, an OL holder
would continue to comply with its OL license to maintain the condition
of the reactor and, by doing so, would safely store the irradiated fuel
in the reactor. If the OL is to be terminated, the OL holder would need
to request and be issued a part 72 specific license to store the
irradiated fuel in a storage installation at the operating site.
Proposed Sec. 57.300(b) would permit the holder of a manufacturing
license under proposed part 57 to store at the manufacturing site the
irradiated fuel from a reactor manufactured under the ML, operated
under the OL, and returned to the manufacturing site. Under this
scenario, the ML holder would need a part 70 license for possession of
the SNM contained in the fuel and a part 72 site-specific license to
allow storage of the irradiated fuel. The ML holder could store the
reactor's irradiated fuel within the reactor if the reactor has been
certified as a part 72 irradiated fuel storage system or move the
reactor's irradiated fuel to another NRC-certified irradiated fuel
storage system. In the cases where the ML holder may temporarily allow
fuel to remain within a reactor, either after operational testing and
before shipment, or when a reactor containing irradiated fuel is
returned to the manufacturing facility site, the ML holder must
demonstrate that the fuel in the reactor is maintained in a safe
condition and that dose to the workers and the public is limited,
consistent with the provisions provided in part 72. Proposed Sec.
57.300(b) would not require the reactor to be a certified storage
system under part 72 because the duration of the storage condition is
expected to be limited as determined by the ML holder's safety
evaluation.
Alternatively, under proposed Sec. 57.300(c), the OL or ML holder
may move the irradiated fuel to another part 72 licensed storage
facility either by transporting the reactor still containing the
irradiated fuel as an NRC-certified transportation package or by
repackaging the irradiated fuel in an NRC-certified transportation
package.
Proposed Sec. 57.300(d), ``Irradiated fuel storage plan,'' would
apply to a holder of a proposed part 57 OL, or a holder of a proposed
part 57 ML that plans to store the irradiated fuel from a reactor
manufactured under the ML, that did not request NRC approval of an
irradiated fuel management and funding plan with its license
application. Such a licensee would be required to submit, for NRC
review and approval under proposed Sec. 57.310, a plan describing how
the licensee intends to manage and provide funding for the management
of all irradiated fuel at a designated storage site following permanent
cessation of operations of the reactor. This submission would need to
occur within 1 year following permanent cessation of reactor
operations, more than 2 years before expiration of the OL if storage
would occur at the operating site, or more than 2 years before the
expiration of the ML if the storage would occur at the manufacturing
site.
2. Decommissioning
Proposed Sec. 57.305, ``Decommissioning and license termination,''
would contain the decommissioning requirements and is generally
consistent with the framework provided in Sec. 50.82(b). The proposed
rule would accommodate the decommissioning of individual microreactors
separate from the overall site, allowing licensees to use the structure
of Sec. 50.82(b)(4), tailored to the design characteristics of the
licensee's facility.
In proposed Sec. 57.60(a)(8)(xvii), applicants would be able to
request NRC approval of a decommissioning plan as part of the joint
application. Early approval of the decommissioning plan would provide
flexibility to support a range of decommissioning strategies, including
decommissioning individual reactors, transporting reactors to a
designated facility, or full-site decommissioning. This approach would
enable licensees to align decommissioning planning with the specific
designs and operational models of their facilities.
Under proposed Sec. 57.305(b), in the absence of an NRC-approved
decommissioning plan, a licensee would be subject to the requirements
of Sec. 50.82(b). Whether at initial licensing or thereafter, the
decommissioning plan

[[Page 23653]]

would need to be prepared using the framework of Sec. 50.82(b)(4),
limited to those provisions applicable to the design characteristics of
the licensed portion of the facility. The licensee's plan would need to
address, as appropriate, transport to a designated facility for final
decommissioning, final decommissioning of individual modules, or final
decommissioning of the entire facility, and would have to ensure
compliance with all applicable safety and environmental requirements.
While licensees under proposed part 57 would not be required to
submit post-shutdown decommissioning activities reports (required for
large LWRs under Sec. 50.82(a)(4)) or license termination plans, they
would be required to provide decommissioning plans under Sec.
50.82(b). The proposed framework is designed to be sufficiently
flexible to address plausible scenarios involving remediation of
radiological contamination and demolition and dismantlement of
radiologically contaminated structures after reactor shutdown and final
demonstration of compliance with the unrestricted release criteria for
residual radioactive material in Sec. 20.1402, ``Radiological criteria
for unrestricted use,'' that may arise during decommissioning. For
example, deployment models may involve one or several nuclear reactors
at a single site, or operational activities could result in significant
radiological contamination that would need to be remediated in order to
meet the unrestricted release criteria. A licensee may request approval
of a decommissioning plan and actions necessary for license termination
prior to permanent cessation of operations, facilitating a streamlined
transition from operations to decommissioning. The decommissioning
plans covering individually licensed reactors are anticipated to have
relatively short decommissioning timelines. Larger or more complex
sites may have extended periods for decommissioning because any
residual radioactivity in the onsite licensed area or environmental
media and from shared systems may be addressed with the last operating
unit at a nuclear plant. Licensees under proposed part 57 would not be
subject to the 60-year decommissioning requirement in Sec. 50.82(a)(3)
but would be required to complete decommissioning without significant
delay. The decommissioning schedules would be approved by the NRC. The
proposed framework supports a graded approach to decommissioning,
tailored to the specific site, design, operational characteristics, and
radiological conditions.
Proposed Sec. 57.305(c)(1) would describe the decommissioning
trust fund requirements and would be equivalent to Sec.
50.82(a)(8)(i). Proposed Sec. 57.305(c)(2)-(3) would describe the
decommissioning cost estimate annual update requirements and would be
equivalent to Sec. 50.82(a)(8)(v)-(vi), respectively.
Proposed Sec. 57.305(d) would prohibit certain decommissioning
activities and would be equivalent to Sec. 50.82(a)(6).
Proposed Sec. 57.305(e) would specify that the entire nuclear
plant must be decommissioned before the final operating license for a
reactor at the site could be terminated.
3. Termination of License
Proposed Sec. 57.305(f) would identify the license termination
requirements as those in Sec. 50.82(b). A licensee would be required
to submit an application for license termination within 2 years
following permanent cessation of operation. Each application for
termination of a license would need to be accompanied or preceded by
the proposed decommissioning plan. The NRC would terminate the license
under the criteria in Sec. 50.82(b)(6). Proposed Sec. 57.305 would
allow for site-specific flexibility in the decommissioning plan to
accommodate various decommissioning strategies for individual reactors
and nuclear plants at which more than one nuclear reactor operated
during the lifetime of the plant, including shared operational areas
and plant systems This approach would ensure that license termination
could be achieved in a manner that would maintain safety and regulatory
compliance while addressing the operational and design-specific needs
of the facility.

I. Subpart H--Maintaining and Revising Licensing Basis Information

The NRC proposes to establish requirements for the maintenance of
licensing basis information in proposed subpart H to part 57.
Proposed Sec. 57.310 would be equivalent to Sec. 50.90,
``Application for amendment of license, construction permit, or early
site permit,'' and would require that a licensee submit an application
to request an amendment to a license. Under proposed part 57, licensees
would be required to include in their applications an analysis of
whether the amendment would involve no significant hazards
consideration, which would be equivalent to the standards in Sec.
50.92, ``Issuance of amendment.'' Proposed Sec. 57.310(e) would
reference Sec. 50.91, ``Notice for public comment; State
consultation,'' for procedures for the Commission to use for notifying
the public and State of the application requesting an amendment for an
OL.
Proposed Sec. 57.312(a) would require a licensee to use Sec.
50.59 for evaluating changes to an FSAR and determining if an amendment
to an OL is required to implement a change to a facility or procedures.
Proposed Sec. 57.312(b) would allow a holder of a part 57 OL that
authorizes operation of a part 57 manufactured reactor to make changes
in the facility or procedures as described in the FSAR (as updated)
without requesting a license amendment if the changes would be the same
as changes approved by amendment to the ML for the manufactured reactor
and other conditions specified in proposed Sec. 57.312(b) were met.
This proposed requirement would prevent license holders and the NRC
from having to duplicate the amendment process for each manufactured
reactor.
Proposed Sec. 57.315, ``Maintenance and submittal of the final
safety analysis, as updated,'' would provide requirements that would be
equivalent to Sec. 50.71(e) for submitting periodic FSAR updates.
Licensees would be required to submit their updated safety analysis
report every 5 years, equivalent to the timeframe for an NPUF as
required by Sec. 50.71(e)(3)(iv).
Proposed Sec. 57.317, ``Updated decommissioning report,'' would be
similar to current Sec. 50.75(f)(1) and would require a construction
permit holder to submit an update to the information required by
proposed Sec. 57.55(i) (i.e., information in the form of a report
indicating how reasonable assurance will be provided that funds will be
available to decommission the facility) before the NRC would issue each
operating license associated with the construction permit. The
operating license holder would be required to submit subsequent updates
to the report every three years beginning within three years after
issuance of the operating license.

J. Subpart I--Transportation Package Design Certification

Under this rulemaking, the NRC proposes to govern transportation of
fissile material or irradiated fuel and associated components through
the provisions of 10 CFR part 71. Part 71 would apply whether the
fueled microreactor or other transportable reactor with a comparable
risk profile would be transported as the packaging plus the approved
contents or only as

[[Page 23654]]

the approved contents in an NRC-certified transportation package.
1. Fueled Reactor as Transportation Package
A fueled reactor could be designated as the transportation package
with the loaded fuel (unirradiated, irradiated, or both) and associated
components as approved contents. To receive a Certificate of Compliance
(CoC) for a transportation package containing fissile or other
radioactive material, an applicant must submit an application to the
NRC and demonstrate that the transportation package design meets the
requirements of 10 CFR part 71. The requirements of Sec. 71.41(a)
stipulate that a transportation package be subjected to tests
prescribed in Sec. Sec. 71.71 and 71.73 in addition to specific Type B
packages being subject to the provisions of Sec. 71.61. The
regulations in Sec. 71.41(a) and (c) allow the NRC to approve
alternatives to the testing requirements provided that those
alternatives are appropriate for the features being considered and
provide an equivalent level of safety, respectively.
The NRC is proposing in Sec. 57.320(a)(1) to provide an option to
allow the use of a previously endorsed or approved risk methodology or
other risk-informed approach in lieu of meeting specific prescriptive
requirements in 10 CFR part 71 if a fueled reactor would be used as the
transportation package. The NRC endorsed a limited use of a risk-
informed methodology for accident conditions specifically for a
transportable microreactor (SECY-24-0062, ``Risk-Informed Methodology
for a Future Transportable TRISO-Based Micro-Reactor Package
Application''). This endorsed risk methodology is an example of one
approach developed only for accident conditions that could be modified
for use as a framework to craft a design certification pathway under
proposed Sec. 57.320(a)(1). This design certification pathway could be
used for both normal and accident conditions with appropriate
justifications, which would allow a package designer to demonstrate the
transportation package meets or exceeds the current level of safety
provided by the part 71 framework.
2. Fueled Reactor as Approved Contents
The NRC proposes two optional considerations for a licensee with
respect to transporting a fueled reactor designated only as approved
contents: (1) design a new transportation package identifying the
fueled reactor as approved contents and submit an application for
review to the NRC for a new part 71 CoC or (2) use an existing
transportation package design with an amended CoC to allow for the
fueled reactor be designated as approved contents. The licensee (ML or
OL) would be designated as the CoC user if they are not responsible for
design authority of the transportation package and thus are not the CoC
holder, or they would be designated as the CoC holder if they are the
responsible design authority and have been issued a CoC by the NRC.

K. Subpart J--Physical Security Requirements

Proposed subpart J would establish the physical protection program
requirements for licensees under proposed part 57 and present a graded
approach to physical protection requirements. If a licensee could meet
the criterion in proposed Sec. 57.60(a)(8)(v)(A)(3), then the
requirement to protect against the DBT of radiological sabotage would
not be applicable. The criterion in proposed Sec. 57.60(a)(8)(v)(A)(3)
would require a licensee to show that potential consequences resulting
from a DBT-initiated event would result in offsite doses below the
values in Sec. 50.34(a)(1)(ii)(D) even if mitigation and recovery
actions, including any operator action, were unavailable or
ineffective. Where the criterion is met, the resulting physical
protection requirements would be those under proposed Sec.
57.60(a)(8)(v)(A)(1)-(2) for protection of SNM and Category 1 and
Category 2 radioactive material, if applicable.
Proposed subpart J would require that an applicant or licensee
establish a physical security program to protect the reactor against
the DBT for radiological sabotage to provide reasonable assurance that
a DBT-initiated event would result in offsite doses below the values in
Sec. 50.34(a)(1)(ii)(D). The elements of this program would include
required intrusion detection and assessment, security communications,
and security response capabilities but would not establish prescriptive
requirements designed to demonstrate that these elements are met.
Proposed subpart J would establish a requirement to coordinate with
local law enforcement and provide sufficient information and training
to personnel who would be relied upon to interdict and neutralize
threats up to and including the design basis threat of radiological
sabotage. Proposed subpart J also would include requirements to
identify target sets, establish and maintain cybersecurity, insider
mitigation, and individual and vehicle search programs and develop
processes to track the performance of the physical protection program.
Section 170D(a) of the AEA permits the Commission to determine
which licensed facilities are part of a class of licensed facilities
for which NRC-conducted force-on-force exercises are appropriate to
assess the ability of a private security force of a licensed facility
to defend against any applicable DBT. Due to the characteristics of
reactors to be licensed under proposed part 57 and the associated
physical security requirements to protect against radiological
sabotage, it would not be appropriate to require force-on-force
exercises to evaluate the performance of these facilities. Therefore,
reactors licensed under proposed part 57 would not be subject to force-
on-force exercises, but these facilities would still have tailored
security requirements and oversight consistent with their relatively
low risk.

L. Subpart K--Categorical Exclusion

As directed by the Commission in the July 28, 2025, Staff
Requirements Memorandum for SECY-24-0046, ``Implementation of the
Fiscal Responsibility Act of 2023 National Environmental Policy Act
Amendments,'' and in accordance with E.O. 14300 section 5(e), the NRC
is proposing for inclusion in subpart K of proposed part 57 a
categorical exclusion from the requirement to prepare an environmental
assessment or environmental impact statement if an application for an
NRC action under proposed part 57 demonstrates that the licensed action
meets the criteria for the categorical exclusion under proposed Sec.
57.350(b). The licensed action could include the siting of multiple
reactors across a region or at one site, and not just a single
microreactor or other reactor with comparable risk profile. For the
reasons described below, the proposed rule includes a determination in
Sec. 57.350(a) that the criteria in Sec. 57.350(b) describe a
category of actions that do not individually or cumulatively have a
significant effect on the human environment as required by 10 CFR
51.22. If the licensed action does not meet the criteria for the
categorical exclusion under proposed Sec. 57.350(b), then the
application would need to include an environmental report in accordance
with part 51.
The criteria to be met for determining the categorical exclusion
applies to a proposed action would include proposed reactor
environmental plant parameter and site parameter envelope values being
compared to values in Table C-1 of appendix C of part 51.

[[Page 23655]]

These proposed reactor values could be derived from the technical
information in a joint application for a CP and associated OL under
proposed subpart C, an ML application under proposed subpart D, or a
standard design approval application under proposed subpart E. The
derived values could then be compared to the appropriate microreactor-
designated Category 1 plant and site parameter envelope values in
NUREG-2249, ``Generic Environmental Impact Statement for Licensing of
New Nuclear Reactors,'' codified in Table C-1 of appendix C of part 51
for demonstrating the appropriateness of a categorical exclusion. In
NUREG-2249, the NRC addresses the impacts of building and operating new
nuclear reactors anywhere in the United States. NUREG-2249 uses a
technology-neutral approach that identifies and analyzes environmental
issues based on plant parameter and site parameter values, common to
building and operating any nuclear reactor for a limited work
authorization, early site permit, construction permit, operating
license, or combined license. Therefore, NUREG-2249 and its findings
can be applied to microreactors and other reactors with comparable risk
profiles under proposed part 57. As such, NUREG-2249 and its findings
can also be applied as the basis for a categorical exclusion for
Category 1 issues, which are issues that the Commission has determined
are SMALL at all sites as long as the proposed action is within the
bound of the relevant values and assumptions in NUREG-2249, and there
is no new and significant information.
For instance, all radiological issues within NUREG-2249 are SMALL
(see

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